Page 1 of 263

DOE Review Release 09/03/2020

Rocky Flats Plant – Occupational Internal 

Dose

ORAUT-TKBS-0011-5 Rev. 04

Effective Date: 09/01/2020

Supersedes: Revision 03

Subject Expert(s): Mutty M. Sharfi, Matthew G. Arno, and David Allen

Document Owner 

Approval: Signature on File

Mutty M. Sharfi, Document Owner

Approval Date: 03/07/2020

Concurrence: Signature on File

Christopher M. Tornes, Objective 1 Representative

Concurrence Date: 03/06/2020

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 03/06/2020

Concurrence: Signature on File

Scott R. Siebert, Project Director Representative

Concurrence Date: 03/06/2020

Approval: Signature on File

Timothy D. Taulbee, Associate Director for Science

Approval Date: 08/06/2020

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 2 of 263

 

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. 

08/17/2007 02 Approved Revision 02 initiated to capture Advisory Board comments 

associated with the June 2007 Advisory Board meeting. 

Incorporation of thorium and internal coworker from ORAUT-OTIB0038 and OCAS-TIB-014. Incorporates formal internal and NIOSH 

review comments. Training required: As determined by the Task 

Manager. Initiated by Mutty M. Sharfi.

09/30/2014 03 Revision initiated to incorporate Advisory Board comments, the 

approval of SEC-00192, and new dose reconstruction approaches in 

assessing tritium, 233U, and recycled uranium. Incorporates 

additional clarifications on current dose reconstruction guidance. 

Incorporates formal internal and NIOSH review comments. 

Constitutes a total rewrite of the document. Training required: As 

determined by the Objective Manager. Initiated by Mutty M. Sharfi.

09/01/2020 04 Revision initiated to capture Advisory Board comments and items 

associated with SEC-00192. Includes update to the co-exposure 

study in Attachment D to capture the data from OCAS-TIB-0014. 

Added the potential and intake rates for type SS plutonium. 

Incorporates formal internal and NIOSH review comments. 

Constitutes a total rewrite of the document. Training required: As 

determined by the Objective Manager. Initiated by Mutty M. Sharfi.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 3 of 263

 

TRADEMARK INFORMATION

ABACOS-PlusTM is a trademark of Canberra Industries in the United States and/or other countries.

All other trademarks are the property of their respective owners.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 4 of 263

 

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................. 14

5.1 Introduction ............................................................................................................................ 17

5.1.1 Purpose...................................................................................................................... 18

5.1.2 Scope......................................................................................................................... 18

5.1.3 Special Exposure Cohort............................................................................................ 18

5.1.4 Federal Bureau of Investigation Raid in 1989 ............................................................. 19

5.2 Source Term .......................................................................................................................... 20

5.2.1 Plutonium ................................................................................................................... 20

5.2.1.1 Isotopic Composition.................................................................................... 20

5.2.1.2 Plutonium Solubility and Particle Size .......................................................... 22

5.2.2 Americium .................................................................................................................. 22

5.2.2.1 Isotopic Composition.................................................................................... 22

5.2.2.2 Americium Solubility and Particle Size ......................................................... 23

5.2.3 Uranium...................................................................................................................... 23

5.2.3.1 Enriched Uranium ........................................................................................ 23

5.2.3.1.1 Isotopic Composition ................................................................. 23

5.2.3.1.2 Enriched Uranium Solubility and Particle Size ........................... 23

5.2.3.2 Depleted Uranium ........................................................................................ 23

5.2.3.2.1 Isotopic Composition ................................................................. 23

5.2.3.2.2 Depleted Uranium Solubility and Particle Size........................... 24

5.2.3.3 Uranium-233 ................................................................................................ 24

5.2.3.4 Recycled Uranium........................................................................................ 24

5.2.4 Thorium ...................................................................................................................... 25

5.2.5 Neptunium.................................................................................................................. 25

5.2.6 Critical Mass Laboratory............................................................................................. 26

5.3 In Vitro ................................................................................................................................... 26

5.3.1 Plutonium Urinalysis ................................................................................................... 26

5.3.1.1 Methods, Units, Isotopes, and Interferences ................................................ 26

5.3.1.2 Plutonium Reporting Levels, Minimum Detectable Amounts, and 

Uncertainties................................................................................................ 27

5.3.2 Americium Urinalysis .................................................................................................. 28

5.3.2.1 Methods, Units, Isotopes, and Interferences ................................................ 28

5.3.2.2 Americium Reporting Levels, Minimum Detectable Amounts, and 

Uncertainties................................................................................................ 28

5.3.3 Uranium Urinalysis ..................................................................................................... 29

5.3.3.1 Enriched Uranium ........................................................................................ 29

5.3.3.1.1 Methods, Units, Isotopes, and Interferences.............................. 29

5.3.3.1.2 Reporting Levels, Minimum Detectable Amounts, and 

Uncertainties ............................................................................. 29

5.3.3.2 Depleted Uranium ........................................................................................ 30

5.3.3.2.1 Methods, Units, Isotopes, and Interferences.............................. 30

5.3.3.2.2 Depleted Uranium Reporting Levels, Minimum Detectable 

Amounts, and Uncertainties ...................................................... 30

5.3.4 Gross Alpha Urinalysis ............................................................................................... 31

5.3.4.1 Methods, Units, Isotopes, and Interferences ................................................ 31

5.3.4.2 Reporting Levels, Minimum Detectable Amounts, and Uncertainties............ 31
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5.3.5 Unmonitored Tritium Exposure ................................................................................... 32

5.3.5.1 Reporting Levels, Minimum Detectable Amounts, and Uncertainties............ 32

5.3.6 Alvin E. Blackwell Bioassay Laboratory ...................................................................... 32

5.4 In Vivo.................................................................................................................................... 32

5.4.1 Americium and Plutonium........................................................................................... 33

5.4.1.1 Methods, Units, Isotopes, and Interferences ................................................ 33

5.4.1.2 Reporting Levels, Minimum Detectable Amounts, and Uncertainties............ 33

5.4.2 Thorium and Depleted Uranium.................................................................................. 34

5.4.2.1 Methods, Units, Isotopes, and Interferences ................................................ 34

5.4.2.2 Reporting Levels, Minimum Detectable Amounts, and Uncertainties............ 36

5.5 Other Bioassay Data.............................................................................................................. 36

5.5.1 Wound Count Data..................................................................................................... 36

5.5.2 Nasal Smears and Fecal Samples.............................................................................. 37

5.6 Records and Reports ............................................................................................................. 37

5.6.1 Urinalysis Records and Reports ................................................................................. 38

5.6.2 Interpretation of the Urinalysis Record Card ............................................................... 38

5.6.3 Interpretation of the Health Sciences Data System – Urinalysis Detail Report............ 38

5.6.4 Interpretation of Other Urinalysis Reports................................................................... 39

5.6.5 Lung Count Records and Reports .............................................................................. 41

5.7 Attributions and Annotations .................................................................................................. 44

References ........................................................................................................................................ 45

Glossary ............................................................................................................................................ 65

ATTACHMENT A MINIMUM DETECTABLE AMOUNTS FOR URINALYSIS .............................. 69

ATTACHMENT B MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS............ 88

ATTACHMENT C EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS .................. 106

ATTACHMENT D INTERNAL CO-EXPOSURE DOSIMETRY DATA......................................... 138

ATTACHMENT E EVALUATION OF THE POTENTIAL FOR INTERNAL DOSE FROM 

NEPTUNIUM-237 ......................................................................................... 204

ATTACHMENT F REPORTED PRESENCE OF MAGNESIUM-THORIUM ALLOY................... 211

ATTACHMENT G TRITIUM ISSUES ......................................................................................... 216

ATTACHMENT H CRITICAL MASS LABORATORY ................................................................. 223

ATTACHMENT I FEDERAL BUREAU OF INVESTIGATION RAID.......................................... 247

LIST OF TABLES

TABLE TITLE PAGE

5-1 Weight percent and fraction of alpha activity for WG plutonium.............................................. 21

5-2 Weight percent and fraction of alpha activity for ZPPR plutonium .......................................... 21

5-3 Weight percent and fraction of alpha activity for EU ............................................................... 23

5-4 Weight percent and fraction of alpha activity for DU............................................................... 24

5-5 Activity fraction of contaminant in recycled uranium ............................................................... 24
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5-6 Median MDA values for plutonium.......................................................................................... 28

5-7 Median MDA values for americium......................................................................................... 29

5-8 Median MDAs for EU .............................................................................................................29

5-9 Median MDAs for DU from 1952 to April 1964........................................................................ 30

5-10 Median MDAs for DU from May 1964 to the present .............................................................. 31

5-11 Median MDAs for gross alpha measurements........................................................................ 31

5-12 Unmonitored tritium exposure rates ....................................................................................... 32

5-13 Summary of MDAs for 241Am.................................................................................................. 35

A-1 Method codes ........................................................................................................................72

A-2 Correlation of method code and analyte................................................................................. 72

A-3 Values of tolerance and reporting levels................................................................................. 72

A-4 Sample volumes for routine 24-hour urine samples................................................................ 77

A-5 Detector background for gas flow proportional counters......................................................... 78

A-6 Median and 95th-percentile blank count rates........................................................................ 78

A-7 Efficiencies of alpha-counting detectors ................................................................................. 79

A-8 Recovery fractions used in MDA assessments....................................................................... 79

A-9 Fractions of alphas absorbed in residue at the 95th percentile............................................... 80

A-10 Gross alpha calibration factor................................................................................................. 81

A-11 Values of variables and MDA for plutonium for median conditions ......................................... 82

A-12 Values of variables and MDA for plutonium for extreme conditions ........................................ 82

A-13 MDA for plutonium for one, two, or three extreme conditions ................................................. 82

A-14 Values of variables and MDA for EU for median conditions.................................................... 83

A-15 Values of variables and MDA for EU for extreme conditions .................................................. 83

A-16 MDA for EU for one, two, or three extreme conditions............................................................ 84

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

extreme conditions.................................................................................................................84

A-18 Values of variables and MDA for americium for median conditions ........................................ 84

A-19 Values of variables and MDA for americium for extreme conditions ....................................... 85

A-20 Values of the MDA for americium for one, two, or three extreme conditions........................... 85

A-21 Values of variables and MDA for gross alpha measurements for median conditions .............. 86

A-22 Values of variables and MDA for gross alpha measurements for extreme conditions............. 86

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

conditions...............................................................................................................................86

B-1 Discontinuity factors...............................................................................................................93

B-2 Calibration factors for the eight-detector Ortec germanium system ........................................ 97

B-3 Calibration factors for the eight-detector PGT I germanium system........................................ 98

B-4 Calibration factors for the eight-detector PGT II germanium system....................................... 99

B-5 Values of variables for the PGT organ pipe germanium detector system ............................. 100

B-6 Calibration factors for the six-detector PGT organ pipe germanium detector system ............ 100

B-7 Values of variables for the EG&G organ pipe germanium detector system .......................... 101

B-8 Calibration factors for the six-detector EG&G organ pipe germanium detector system......... 101

B-9 MDA conversion factors for values of ppm 241Am ................................................................. 102

B-10 Americium-241 ingrowth in plutonium................................................................................... 102

B-11 Americium-241 MDA for in vivo lung counts......................................................................... 104

D-1 Summary of uranium urinary excretion rate analyses, 1953 to 2005 .................................... 151

D-2 Summary of plutonium urinary excretion rate analyses, 1952 to 2005.................................. 153

D-3 Americium-241 lung count bioassay data............................................................................. 156

D-4 Derived uranium intake rates (95th-percentile), type F material, 1953 to 2005..................... 157

D-5 Derived uranium intake rates (95th-percentile), type M material, 1953 to 2005 .................... 157

D-6 Derived uranium intake rates (95th-percentile), type S material, 1953 to 2005..................... 157

D-7 Derived type M plutonium intake rates (95th-percentile), 1952 to 2005 ................................ 158

D-8 Derived type S 239Pu intake rates (95th-percentile), 1952 to 2005........................................ 158

D-9 Derived type SS 239Pu intake rates (95th-percentile), 1952 to 2005...................................... 158
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E-1 Neptunium-237 fiscal-year-end inventories .......................................................................... 207

E-2 Mean masses and ratios of 239Pu and 237Np in waste drums ................................................ 208

E-3 Mean mass concentrations and ratios of 239Pu and 237Np in RFP waste drums .................... 208

E-4 Neptunium receipts from other sites..................................................................................... 209

E-5 Neptunium shipments from RFP to receiving sites ............................................................... 210

G-1 Reported tritium air concentrations from the August 30, 1974, release................................. 219

H-1 Average reactor power and fission rates derived from measurements at CML ..................... 228

H-2 Maximum removable alpha contamination ........................................................................... 229

H-3 Maximum dose rates in CML experimental and material storage areas................................ 229

H-4 Room 101 walk-in hood contamination results, November 1989 to April 1990 ..................... 233

H-5 Air sample results summary for the CML Building 886 nuclear materials storage and 

experimental areas............................................................................................................... 238

H-6 Experiment periods, contamination incident dates, and associated records of airborne 

contamination results ........................................................................................................... 239

H-7 HEU solution spills at CML................................................................................................... 243

H-8 Accumulated activities of dosimetrically significant MFAPs in UNH fuel at the CML............. 244

H-9 Maximum committed organ doses from inhalation of airborne MFAPs at CML..................... 245

I-1 Available internal sample data, 1980 to 1995....................................................................... 260

LIST OF FIGURES

FIGURE TITLE PAGE

C-1 Urinalysis Record Card and HSDS – Urinalysis Detail (1) (first activity date on the HSDS 

portion 10-29-54) ................................................................................................................. 107

C-2 Urinalysis Record Card and HSDS – Urinalysis Detail (2) (first activity date on the HSDS 

portion 8-19-53) ................................................................................................................... 108

C-3 Urinalysis Record Card and HSDS – Urinalysis Detail (3) (first activity date on the HSDS 

portion: 1-6-58) ................................................................................................................... 109

C-4 HSDS – Urinalysis Detail (1) (first activity date 9-17-58) ...................................................... 110

C-5 HSDS – Urinalysis Detail (2) (first activity date 3-19-73) ...................................................... 111

C-6 Analytical Report – Bioassay Analysis Data 3-15-93............................................................ 112

C-7 Analytical Report – Bioassay Analysis Data 10-28-93 .......................................................... 113

C-8 Form 1 – Sample Results 1-29-96........................................................................................ 114

C-9 Rocky Flats Environmental Technology Site (1) 8-27-96...................................................... 115

C-10 Rocky Flats Environmental Technology Site (1) 8-8-96........................................................ 116

C-11 Form 1 – Sample Results – Quanterra, Richland 7-31-98 .................................................... 117

C-12 General Engineering Laboratories 6-28-99 .......................................................................... 118

C-13 Health Sciences Urinalysis Record (with tritium, fecal, and nasal smear results) ................. 119

C-14 Health Physics – Body Counter Information 12-8-65............................................................ 120

C-15 Health Physics – Body Counter Information 5-16-68............................................................ 121

C-16 Health Physics – Body Counter Information 8-26-68............................................................ 122

C-17 Health Physics – Body Counter Information 9-16-70............................................................ 124

C-18 Radiation Dosimetry – Body Count Results 10-3-74............................................................. 124

C-19 Radiation Dosimetry – Body Count Results 5-30-75............................................................. 125

C-20 Radiation Dosimetry – Body Count Results 1-9-78............................................................... 127

C-21 Body Counter Results 12-8-81 ............................................................................................. 128

C-22 Radiation Dose Assessment – Body Count Results 7-22-83................................................ 128

C-23 Radiation Dose Assessment – Body Count Results 5-18-83................................................ 130

C-24 Radiation Dose Assessment – Body Count Results 2-21-84................................................ 131

C-25 Radiation Dose Assessment – Body Count Results 3-22-84................................................ 132

C-26 Radiation Dose Assessment – Body Count Results 10-10-85 .............................................. 133
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C-27 Radiation Dose Assessment – Body Count Results 3-6-89.................................................. 134

C-28 Internal Dosimetry – Lung Count Results 11-23-93.............................................................. 135

C-29 ABACOS-Plus 3-6-96........................................................................................................... 135

C-30 ABACOS-Plus 11-15-01....................................................................................................... 137

C-31 ABACOS-Plus 6-14-01......................................................................................................... 137

D-1 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1976, 50thpercentile, type F ................................................................................................................. 159

D-2 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 50thpercentile, type F ................................................................................................................. 159

D-3 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1989 to 12/31/1993, 

50th-percentile, type F ......................................................................................................... 160

D-4 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1994 to 12/31/2005, 

50th-percentile, type F ......................................................................................................... 160

D-5 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/2005, 50th-percentile, type F ...................................................................................... 161

D-6 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1976, 84thpercentile, type F ................................................................................................................. 161

D-7 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 84thpercentile, type F. ................................................................................................................ 162

D-8 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1989 to 12/31/1993, 

84th-percentile, type F ......................................................................................................... 162

D-9 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1994 to 12/31/2005, 

84th-percentile, type F ......................................................................................................... 163

D-10 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/2005, 84th-percentile, type F ...................................................................................... 163

D-11 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1976, 50thpercentile, type M ................................................................................................................. 164

D-12 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 50thpercentile, type M ................................................................................................................. 164

D-13 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1989 to 12/31/1993, 

50th-percentile, type M ......................................................................................................... 165

D-14 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1994 to 12/31/2005, 

50th-percentile, type M ......................................................................................................... 165

D-15 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/2005, 50th-percentile, type M ..................................................................................... 166

D-16 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1988, 

84th-percentile, type M ......................................................................................................... 166
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D-17 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1989 to 12/31/1993, 

84th-percentile, type M ......................................................................................................... 167

D-18 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1994 to 12/31/2005, 

84th-percentile, type M ......................................................................................................... 167

D-19 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/2005, 84th-percentile, type M ..................................................................................... 168

D-20 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1958, 50thpercentile, type S ................................................................................................................. 168

D-21 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1959 to 12/31/1960, 50thpercentile, type S ................................................................................................................. 169

D-22 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1961 to 12/31/1963, 50thpercentile, type S ................................................................................................................. 169

D-23 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1964 to 12/31/1976, 50thpercentile, type S ................................................................................................................. 170

D-24 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 50thpercentile, type S ................................................................................................................. 170

D-25 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1989 to 12/31/1993, 50thpercentile, type S ................................................................................................................. 171

D-26 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/19e94 to 12/31/2005, 50thpercentile, type S ................................................................................................................. 171

D-27 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/2005, 50th-percentile, type S ...................................................................................... 172

D-28 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1958, 84thpercentile, type S ................................................................................................................. 172

D-29 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1959 to 12/31/1960, 84thpercentile, type S ................................................................................................................. 173

D-30 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1961 to 12/31/1963, 84thpercentile, type S ................................................................................................................. 173

D-31 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1964 to 12/31/1976, 84thpercentile, type S ................................................................................................................. 174

D-32 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 84thpercentile, type S ................................................................................................................. 174

D-33 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1989 to 12/31/1993, 84thpercentile, type S ................................................................................................................. 175
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 10 of 263

 

D-34 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1994 to 12/31/2005, 84thpercentile, type S ................................................................................................................. 175

D-35 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/2005, 84th-percentile, type S ...................................................................................... 176

D-36 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 50thpercentile, type M ................................................................................................................. 176

D-37 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1969, 50thpercentile, type M ................................................................................................................. 177

D-38 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake

rates compared with measured plutonium-in-urine results, 1/1/1970 to 12/31/1979, 50thpercentile, type M ................................................................................................................. 177

D-39 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1980 to 12/31/1988, 50thpercentile, type M ................................................................................................................. 178

D-40 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1989 to 12/31/1993, 50thpercentile, type M................................................................................................................. 178

D-41 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1994 to 12/31/2005, 50thpercentile, type M................................................................................................................. 179

D-42 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, from all intakes 1/1/1952 to 

12/31/1988, 50th-percentile, type M ..................................................................................... 179

D-43 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 84thpercentile, type M ................................................................................................................. 180

D-44 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1969, 84thpercentile, type M ................................................................................................................. 180

D-45 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1970 to 12/31/1979, 84thpercentile, type M ................................................................................................................. 181

D-46 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1980 to 12/31/1988, 84thpercentile, type M ................................................................................................................. 181

D-47 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1989 to 12/31/1993, 84thpercentile, type M ................................................................................................................. 182

D-48 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1994 to 12/31/2005, 84thpercentile, type M ................................................................................................................. 182

D-49 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, from all intakes 1/1/1952 to 

12/31/2005, 84th-percentile, type M ..................................................................................... 183

D-50 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 50thpercentile, type S ................................................................................................................. 183
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D-51 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1971, 50thpercentile, type S ................................................................................................................. 184

D-52 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1972 to 12/31/1979, 50thpercentile, type S ................................................................................................................. 184

D-53 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1980 to 12/31/1993, 50thpercentile, type S ................................................................................................................. 185

D-54 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1994 to 12/31/2005, 50thpercentile, type S ................................................................................................................. 185

D-55 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/2005, 50thpercentile, type S ................................................................................................................. 186

D-56 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 84thpercentile, type S ................................................................................................................. 186

D-57 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1971, 84thpercentile, type S ................................................................................................................. 187

D-58 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1972 to 12/31/1979, 84thpercentile, type S ................................................................................................................. 187

D-59 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1980 to 12/31/1993, 84thpercentile, type S ................................................................................................................. 188

D-60 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1994 to 12/31/2005, 84thpercentile, type S ................................................................................................................. 188

D-61 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/2005, 84thpercentile, type S ................................................................................................................. 189

D-62 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1972 to 12/31/1976, 

50th-percentile, type S ......................................................................................................... 189

D-63 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1977 to 12/31/1982, 

50th-percentile, type S ......................................................................................................... 190

D-64 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1983 to 12/31/1988, 

50th-percentile, type S ......................................................................................................... 190

D-65 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1972 to 12/31/1988, 

50th-percentile, type S ......................................................................................................... 191

D-66 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1972 to 12/31/1976, 

84th-percentile, type S ......................................................................................................... 191

D-67 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1977 to 12/31/1982, 

84th-percentile, type S ......................................................................................................... 192
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D-68 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1983 to 12/31/1988, 

84th-percentile, type S ......................................................................................................... 192

D-69 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1972 to 12/31/1988, 

84th-percentile, type S ......................................................................................................... 193

D-70 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 

12/31/1961, 50th-percentile, type SS ................................................................................... 193

D-71 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1962 to 

12/31/1971, 50th-percentile, type SS ................................................................................... 194

D-72 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1972 to 

12/31/1979, 50th-percentile, type SS ................................................................................... 194

D-73 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 

12/31/1993, 50th-percentile, type SS ................................................................................... 195

D-74 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1994 to 

12/31/2005, 50th-percentile, type SS ................................................................................... 195

D-75 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 

12/31/2005, 50th-percentile, type SS ................................................................................... 196

D-76 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 

12/31/1961, 84th-percentile, type SS ................................................................................... 196

D-77 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1962 to 

12/31/1971, 84th-percentile, type SS ................................................................................... 197

D-78 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1972 to 

12/31/1979, 84th-percentile, type SS ................................................................................... 197

D-79 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 

12/31/1993, 84th-percentile, type SSe ................................................................................. 198

D-80 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1994 to 

12/31/2005, 84th-percentile, type SS ................................................................................... 198

D-81 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 

12/31/2005, 84th-percentile, type SS ................................................................................... 199

D-82 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1972 to 

12/31/1976, 50th-percentile, type SS ................................................................................... 199

D-83 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1977 to 

12/31/1982, 50th-percentile, type SS ................................................................................... 200

D-84 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1983 to 

12/31/1988, 50th-percentile, type SS ................................................................................... 200
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D-85 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots) from all 

intakes 1/1/1972 to 12/31/1988, 50th-percentile, type SS .................................................... 201

D-86 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1972 to 

12/31/1976, 84th-percentile, type SS ................................................................................... 201

D-87 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1977 to 

12/31/1982, 84th-percentile, type SS ................................................................................... 202

D-88 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1983 to 

12/31/1988, 84th-percentile, type SS ................................................................................... 202

D-89 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots) from all 

intakes 1/1/1972 to 12/31/1988, 84th-percentile, type SS .................................................... 203

H-1 Weekly averages of removable alpha contamination in the Building 886 experimental 

and material storage areas, first quarter 1989...................................................................... 230

H-2 Average removable alpha contamination in the Building 886 experimental and material 

storage areas, second quarter 1989..................................................................................... 230

H-3 Weekly averages of removable alpha contamination in the Building 886 offices and 

control points (hallways and conference rooms), first quarter 1989...................................... 231

H-4 Average removable alpha contamination in the Building 886 office area, January to June 

1989..................................................................................................................................... 231

H-5 Floor plan of the assembly room (Room 101) showing the walk-in hood containing the 

vertical split table (V) and solution base (S) experimental locations ..................................... 234

H-6 Memorandum on beta/gamma penetrating and removable surface contamination 

measurements made upon discovering three sealed 90Sr sources in the Building 886 

material storage area ........................................................................................................... 235

H-7 Air sampler locations in the Building 886 cluster .................................................................. 237

I-1 Percent of workers badged for dosimetry ............................................................................. 260
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ACRONYMS AND ABBREVIATIONS 

ABRWH Advisory Board on Radiation and Worker Health

AEB Alvin E. Blackwell Bioassay Laboratory

AEC U.S. Atomic Energy Commission

AMAD activity median aerodynamic diameter

AWE atomic weapons employer

Bq becquerel

CAM continuous air monitor

CEDR Comprehensive Epidemiology Data Resource

CFR Code of Federal Regulations

Ci curie

cm centimeter

CML Critical Mass Laboratory

cpm counts per minute

CWT chest wall thickness

d day

DCF dose conversion factors

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

DTPA diethylene triamine pentaacetic acid

DU depleted uranium

EDTA ethylene diamine tetraacetic acid

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EMCBC Environmental Management Consolidated Business Center

EPA U.S. Environmental Protection Agency

ERDA U.S. Energy Research and Development Administration

EU enriched uranium

F fast (absorption type)

FBI Federal Bureau of Investigation

FOIA Freedom of Information Act

FY fiscal year

g gram

GSD geometric standard deviation

HEU highly enriched uranium

hr hour

HSDS Health Sciences Data System

HT elemental tritium (tritiated gas)

HTO tritiated water vapor

ICRP International Commission on Radiological Protection

IDOT Internal Dosimetry Tool

IMBA Integrated Modules for Bioassay Analysis

in. inch

INL Idaho National Laboratory
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IREP Interactive RadioEpidemiological Program

keV kilovolts-electron, 1,000 electron volts

kg kilogram

L liter

L X-ray low-energy X-ray

LANL Los Alamos National Laboratory

LLNL Lawrence Livermore National Laboratory

M medium (absorption type)

MDA minimum detectable amount

MDL minimum detection level

MFAP mixed fission and activation product

MgTh magnesium-thorium (alloy)

min minute

mL milliliter

MLT minutes live time

mm millimeter

MSS multispectral scanner survey

MPLB maximum permissible lung burden

mR millirad

mrem millirem

mW milliwatt

MW megawatt

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NMMSS Nuclear Materials Management and Safeguards System

NOCTS NIOSH-Division of Compensation Analysis and Support Claims Tracking System

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

Oy bare HEU metal

pCi picocurie

PER Program Evaluation Report

PGT Princeton Gamma Tech

PHA pulse height analysis

ppm parts per million

PuNp plutonium-neptunium (alloy)

QA quality assurance

QC quality control

RCG radioactivity concentration guide

RCRA Resource Conservation and Recovery Act of 1976

RFAO Rocky Flats Area Office

RFP Rocky Flats Plant

ROI region of interest

RWMC Radioactive Waste Management Complex

s second
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S slow (absorption type)

SAAM selective alpha air monitor

SEC Special Exposure Cohort

SNL Sandia National Laboratories

SRDB Ref ID Site Research Database Reference Identification (number)

SRS Savannah River Site

SS super S (absorption type)

SSR safe, secure railcar

SST safe, secure transport

Sv sievert

TBD technical basis document

TBP tributyl phosphate

TLD thermoluminescent dosimeter

TOPO trioctylphosphene oxide

TSA Technical Safety Appraisal

TTA thenoyltrifluoroacetone

U.S.C. United States Code

UNH uranyl nitrate

W watt

WG weapons grade

wk week

WISC Waste Stream Identification and Characterization

yr year

ZPPR Zero Power Plutonium (later Physics) Reactor

μCi microcurie

μg microgram

μm micrometer

γ gamma particle

§ section or sections
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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 historical background information and guidance to assist in the preparation of 

dose reconstructions at particular Department of Energy (DOE) or Atomic Weapons Employer (AWE) 

facilities or categories of DOE or AWE facilities. They will be revised in the event additional relevant 

information is obtained about the affected DOE or AWE facility(ies). These documents may be used 

to assist NIOSH staff in the evaluation of Special Exposure Cohort (SEC) petitions and the completion 

of the individual work required for each dose reconstruction.

In this document the word “facility” is used to refer to an area, building, or group of buildings that 

served a specific purpose at a DOE or AWE facility. It does not mean nor should it be equated to an 

“AWE facility” or a “DOE facility.” The terms AWE and DOE facility are defined in sections 7384l(5) 

and (12) of the Energy Employees Occupational Illness Compensation Program Act of 2000 

(EEOICPA), respectively. An AWE facility means “a facility, owned by an atomic weapons employer, 

that is or was used to process or produce, for use by the United States, material that emitted radiation 

and was used in the production of an atomic weapon, excluding uranium mining or milling.” 42 U.S.C. 

§ 7384l(5). On the other hand, a DOE facility is defined 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 [DOE] (except for buildings, 

structures, premises, grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program);” 

and with regard to which DOE has or had a proprietary interest, or “entered into a contract with an 

entity to provide management and operation, management and integration, environmental 

remediation services, construction, or maintenance services.” 42 U.S.C. § 7384l(12). The Department 

of Energy (DOE) determines whether a site meets the statutory definition of an AWE facility and the 

Department of Labor (DOL) determines if a site is a DOE facility and, if it is, designates it as such.

Accordingly, a Part B claim for benefits must be based on an energy employee’s eligible employment 

and occupational radiation exposure at a DOE or AWE facility during the facility’s designated time 

period and location (i.e., covered employee). After DOL determines that a claim meets the eligibility 

requirements under EEOICPA, DOL transmits the claim to NIOSH for a dose reconstruction. 

EEOICPA provides, among other things, guidance on eligible employment and the types of radiation 

exposure to be included in an individual dose reconstruction. Under EEOICPA, eligible employment 

at a DOE facility includes individuals who are or were employed by DOE and its predecessor 

agencies, as well as their contractors and subcontractors at the facility. Unlike the abovementioned 

statutory provisions on DOE facility definitions that contain specific descriptions or exclusions on 

facility designation, the statutory provision governing types of exposure to be included in dose 

reconstructions for DOE covered employees only requires that such exposures be incurred in the 

performance of duty. As such, NIOSH broadly construes radiation exposures incurred in the 

performance of duty to include all radiation exposures received as a condition of employment at 

covered DOE facilities in its dose reconstructions for covered employees. For covered employees at 

DOE facilities, individual dose reconstructions may also include radiation exposures related to the 

Naval Nuclear Propulsion Program at DOE facilities, if applicable. No efforts are made to determine 

the eligibility of any fraction of total measured exposure for inclusion in dose reconstruction.

NIOSH does not consider the following types of exposure as those incurred in the performance of 

duty as a condition of employment at a DOE facility. Therefore these exposures are not included in 

dose reconstructions for covered employees (NIOSH 2010):

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

conventional structures

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

reasons
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5.1.1 Purpose

This technical basis document (TBD) discusses Rocky Flats Plant (RFP) internal dosimetry data for 

dose reconstruction and includes guidance for the appropriate use of that information.

5.1.2 Scope

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.

Sections 5.3 and 5.4 discuss these two datasets 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 5.6 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 5.7. 

Attachments A and B provide detailed information about minimum detectable amounts (MDAs) for 

urinalysis and in vivo lung counts, respectively. Attachment C provides examples of bioassay data 

records and reports, and Attachment D details internal co-exposure dosimetry data. Attachment E 

evaluates the potential for internal dose from 237Np. Attachment F addresses worker statements 

about the presence of MgTh alloy at RFP, and Attachment G examines tritium issues. Attachment H 

provides additional information on the Critical Mass Laboratory (CML), and Attachment I provides an 

analysis of the potential effects on dose reconstruction due to a Federal Bureau of Investigation (FBI) 

raid and investigation of RFP activities in 1989 (see Section 5.1.4).

5.1.3 Special Exposure Cohort

The Secretary of the U.S. Department of Health and Human Services has designated the following 

class of employees from RFP as an addition to the SEC (Sebelius 2013, p. 3):

All employees of the Department of Energy, its predecessor agencies, and their 

contractors and subcontractors who worked at the Rocky Flats Plant in Golden, 

Colorado, from April 1, 1952, through December 31, 1983, for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment or in 

combination with work days within the parameters established for one or more other 

classes of employees included in the Special Exposure Cohort.

NIOSH has determined that doses to unmonitored RFP workers from neptunium, thorium, and 233U 

(and its associated 232U and 228Th progeny) cannot be reconstructed from April 1, 1952, through 

December 31, 1983, inclusive (NIOSH 2013a). 

The class includes all workers during the SEC period. Because of the identified dose reconstruction 

infeasibility, all dose reconstructions for monitored workers during the SEC period are considered 

partial dose reconstructions. If monitoring data are available for workers in the SEC, dose is to be 

assigned as appropriate based on that data. However, such dose reconstructions are still considered 
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partial dose reconstructions because of the determination that exposure to neptunium, thorium, and 

233U (and its associated 232U and 228Th progeny) during the SEC period cannot be bounded.

5.1.4 Federal Bureau of Investigation Raid in 1989

On October 24 and 25, 1989, a former RFP worker made allegations in a set of interviews conducted 

by the U.S. Environmental Protection Agency (EPA) National Enforcement Investigation Center’s 

Office of Criminal Investigations and the FBI. These interviews resulted from a phone call by the 

interviewee to the FBI Rocky Flats Hotline on June 16, 1989, alleging safety violations and 

manipulation of laboratory samples at RFP.

Attachment I provides a detailed assessment of these allegations. The following is a summary of the 

conclusions of this assessment:

• No scientific basis for concluding that the issues raised about environmental samples would 

compromise radiological count results, nor does the reviewed information corroborate a link 

between the environmental and occupational radiological programs.

• There were no situations identified where falsification or invalidation of data would impact the 

ability to perform dose reconstruction under EEOICPA.

• The charges against Rockwell were specific to environmental Resource Conservation and 

Recovery Act of 1976 (RCRA) and Clean Water Act laws and the impact to the environment; 

the charges did not specifically call out a data falsification, data validity issues, or a data 

quality violation.

• One individual provided information about involvement in shredding documents. While the 

documents being destroyed could have been some kind of field surveys, there is no indication 

that those surveys have an impact on the ability to bound or reconstruct dose for the class, as 

long as the personnel monitoring data exist. These records do exist in the associated 

personnel files in the NIOSH-Division of Compensation Analysis and Support Claims Tracking 

System (NOCTS); thus, those files were not destroyed.

• An issue of “penciling-in” information on radiological field survey records was raised. The 

primary source of radiological information for individual dose reconstruction is the individual 

thermoluminescent dosimeter (TLD) results and bioassay information. These are performed in 

a laboratory and not documented in the field. 

• Concerns about bioassay sample analysis results (false positives and [statistical] variations); 

bioassay sample handling and processing; personnel contamination and contamination 

incidents; and issues about tritium bubblers, neptunium, MgTh alloy, and the CML. The dose 

reconstruction process accounts for the potential for missed doses and incorporates methods 

that are favorable to the claimant. All of the issues concerning tritium bubblers, neptunium, 

MgTh Alloy, and the CML are addressed in other sections of this TBD. 

• A contention was made that there was an additional August 1989 aerial multispectral scanner 

survey (MSS) performed at RFP in addition to the one performed in June and July of 1989, 

and that the flyover data indicate the presence of the isotopes 137Cs and 90Sr, which is used to 

imply that an unreported criticality occurred. Neither the August 1989 flyover survey nor 

evidence supporting a criticality event could be located. Based on interviews, document 

reviews, and files provided, no evidence or information was identified that disputes the ability 

to bound RFP worker dose under EEOICPA.
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• A review of the RFP Technical Safety Appraisal (TSA) concluded that there was no identified 

impact on radiological personnel monitoring data that are used to support 

bounding/reconstructing dose for the RFP worker class.

• A review of an allegation of record falsification involving mislabeling waste for shipment off site

concluded that none of this information impacts the ability to reconstruct radiation dose with

sufficient accuracy under the EEOICPA Program for the RFP worker class.

• A review of excerpts from the notebook/logbook of an RFP Manager who made notes on the 

radiological program did not support a data falsification issue that would impact the ability to

reconstruct dose for the RFP worker class.

• A review of the quantity of available personnel radiological monitoring data available at the 

time of this assessment. It was concluded that there exists a sufficient quantity of individual 

monitoring data to support the assessment of RFP personnel doses.

Based on this information, the conclusion was reached that there was no impact to the ability to 

perform individual dose reconstructions with sufficient accuracy under EEOICPA due to the FBI raid 

or issues about data falsification or invalidation.

5.2 SOURCE TERM

In 1993, the Secretary of Energy formally announced the end of nuclear production at RFP. 

Remediation was completed at the RFP in late 2005. Co-exposure intakes should be assigned, when 

applicable, up through 2005. Only environmental intakes should be assigned after 2005.

5.2.1 Plutonium

5.2.1.1 Isotopic Composition

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

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

count data; 

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

account for the dose from unmeasured isotopes; and

• 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.

Table 5-1 lists the weight percent and fraction of alpha activity for each isotope of weapons-grade 

(WG) plutonium that was present at RFP throughout most of its 1952-to-1989 production history.

The Zero Power Plutonium (later 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 that involved ZPPR plutonium generally note “ZPPR” or “ZPPR material,” especially on the 

lung count reports (RFP 1976–1996, pp. 99, 225). 
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Table 5-1. Weight percent and fraction of alpha activity for WG 

plutonium.

a

Isotope Weight percent

Fraction of

alpha activityb

Pu-238 0.01 0.023

Pu-239 93.79 0.8

Pu-240 5.8 0.18

Pu-241 0.36b Not applicable

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. 236). 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 (KHC 2002, p. 120).

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 Not applicable

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

undocumented source.

Dose reconstructions 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 involved waste or byproduct (separated 241Am) from the purification of aged 

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

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

for involved workers. A nominal amount, 100 or 1,000 ppm by mass, of 241Am should be assumed if 

no other data are available. Note that the practice at RFP was to measure the ppm 241Am in a 

representative sample of material that was 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. The fact that RFP arbitrarily assumed 

1,000 ppm should not be the basis for determining the plutonium mixture.

If the plutonium intake for WG plutonium is assessed for 239,240Pu, the activity of 241Am in the intake 

mixture is calculated by: 

241Am activity = 239,240Pu activity × [48.2 × ppm 241Am ÷ (1 × 106

 – ppm 241Am)].

For ZPPR plutonium: 

241Am activity = 239,240Pu activity × [44.6 × ppm 241Am ÷ (1 × 106

 – ppm 241Am)].

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 in the ZPPR multiplying factor.

In December 1989, DOE suspended plutonium processing. Therefore, RFP would not have received 

any new plutonium shipments after this date. However, the americium in the residual plutonium 
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mixture would have continued to age over time. To account for this aged material, after 1989, only 

the 6% 10-year-aged plutonium mixture ratio should be considered, based on guidance in the 

ORAUT-TKBS-0006-5, Hanford Site – Occupational Internal Dose (ORAUT 2015d). The Hanford 

data is being used as a surrogate because it was the main source of RFP’s plutonium material. This 

results in a 239+240Pu:238Pu:

241Pu:241Am of 1.00:0.111:7.150:0.146 and a 241Pu:241Am activity ratio of 

49.12:1 (4,912%). 

5.2.1.2 Plutonium Solubility and Particle Size

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 (type SS) characteristics (ORAUT 2020). 

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

insoluble (type SS), or a mixture of solubilities. Dose reconstructors should select the material type 

that is most favorable to the claimant (ORAUT 2018). 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 (Dow 1965), 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. Therefore, for individuals potentially involved with a plutonium fire, the 

more favorable particle size of 1-μm or 5-μm AMAD should be assumed. 

The 1-μm particle adjustment for RFP plutonium fires should only be applied for energy employees 

who were involved with a known intake from a plutonium fire (or any time dose reconstructors deem 

use of a 1-μm AMAD particle size appropriate) (NIOSH 2002). This can be from involvement with the 

plutonium fire itself, including being in the building or area and exposed to smoke or airborne activity 

from the fire as well as involvement in cleanup activities immediately after the fire.

The application of the 1-μm particle size adjustment only applies to individuals who were involved in a 

fire (i.e., operators and firefighters) and the individuals who performed the immediate cleanup of the 

incident. Once that is accomplished, it is assumed that the particle size reverts back to the default 

5-μm AMAD. When applicable, the adjustment factor is applied only to the dose associated with the 

intake that is directly from the fire and cleanup. The 1-μm particle size adjustment typically applies for 

a short period (i.e., days, weeks, or a few months).

The use of the 1-μm particle adjustment for RFP plutonium fires is specific to the intake being 

assessed. If an earlier or later intake is assessed that is not associated with a plutonium fire, the 

1-μm particle adjustment factor does not apply.

5.2.2 Americium

5.2.2.1 Isotopic Composition

For the NIOSH Dose Reconstruction Project, the measured americium is 241Am. The source of the 

americium is only from the decay of 241Pu. No other americium isotopes are involved.
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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 that are 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 1995, p. 85). For 

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

solubility classification for the plutonium particle in Section 5.2.1.2 (ICRP 1994).

Dose reconstructors should use the default 5-μm AMAD particle size (NIOSH 2002) except for fire 

incidents, in which a 1-μm AMAD should be assumed for consistency with Section 5.2.1.2 above.

5.2.3 Uranium

5.2.3.1 Enriched Uranium

5.2.3.1.1 Isotopic Composition

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 EU.

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

a. Source: DOE (1980, Volume 1, Table 2.7.2-4, p. 238).

5.2.3.1.2 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 or UNH) to inhalation type F; 

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

1979, 1994, 1995). All of these compounds were involved in the recovery and recycling processes for 

EU in Building 881 (KHC 2000a).

In many cases, the compound of uranium in an intake was 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).

5.2.3.2 Depleted Uranium

5.2.3.2.1 Isotopic Composition

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 

(Table 5-4).
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Table 5-4. Weight percent and fraction of alpha activity for DU.

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. 238).

5.2.3.2.2 Depleted Uranium Solubility and Particle Size

Operations with DU involved metalworking including casting, forming, and melting with what was 

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

between type S and type M (KHC 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).

5.2.3.3 Uranium-233

Operations with 233U (thorium strikes) occurred between 1964 and 1983 (Moment, Gibbs, and 

Freiboth 1999). The process included the following steps:

1. Material received as nitrate solution,

2. Thorium strike (thorium fluoride precipitation > peroxide precipitation > UO4 cake),

3. Conversion (UO4 > UO3 > UO2 > UF4),

4. Reduction to metal and casting into an ingot,

5. Rolling ingot into a sheet and producing part blanks from the sheet,

6. Machining, and

7. Sampling.

In the beginning of operations, the first two steps were performed in Building 71 (later called 771). 

Intermediate steps (conversion to UF4, reduction to metal, and casting) were performed in Building 81 

(later called 881). The ingot was rolled and formed into parts in Building 83 (later called 883) and then 

transferred back to Building 81 for final machining. Finished parts were sent to Building 77 (later 

called 777) where they were assembled and shipped. By the mid-1970s, the intermediate steps in 

Building 881 shifted to the research and development areas of Building 771.

Because of data issues and limitations, no specific methods to bound doses from 233

U and 232U have 

been determined. Therefore, doses to unmonitored RFP workers from neptunium, thorium, and 233U 

(and its associated 232U and 228Th contaminants) cannot be reconstructed.

5.2.3.4 Recycled Uranium

For all DOE uranium after 1952, this analysis assumed the possibility that uranium from refineries was 

recycled uranium or contained recycled uranium. Table 5-5 provides the activity fractions that should 

be applied to all uranium intakes after 1952 (NIOSH 2011).

Table 5-5. Activity fraction of contaminant in recycled uranium. 

Recycled uranium contaminant Pu-239 Np-237 Tc-99 Th-232 Th-228

Activity fraction of contaminant in uranium 0.00246 0.00182 0.379 2.73E-06 2.73E-06
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Note: If plutonium intakes are assigned through bioassay or co-exposure, it should not be 

assigned as part of recycled uranium, as this would result in a double assignment of 

plutonium. However, the rest of the contaminants should still be assigned.

5.2.4 Thorium

Thorium was present at RFP facilities from the beginning of operations in 1952 at least through 1975; 

quantities varied from 0 or gram quantities to 238 kg in any particular month at the site (ChemRisk 

1992, p. 136; Ulsh et al. 2008; Author unknown 1976a). The site used thorium in various ways 

including:

• Fabrication of metal parts from natural thorium or thorium alloys,

• Use of oxide (“thoria”) as a mold-coating compound,

• In compounds for numerous analytical procedures and research and development programs,

• As a substitute for uranium or plutonium components in various research and development 

activities and programs, and

• The removal of 228Th (thorium strike) performed during 233U processing.

While the consensus of the contributors and authors of the thorium reference documents was that the 

quantities and concentrations of thorium on the site over the years at RFP were minimal, there was 

the potential for thorium exposures to certain populations of workers. The available documentation 

supports the presence of thorium on site in the early 1950s through the development of internal and 

external thorium-monitoring processes (Dow 1953–1963; Hammond 1956–1958, 1958).

There is no indication that the MgTh was ever used at the RFP site. Attachment F presents an 

analysis of available information to support this conclusion.

Because of data issues and limitations, no specific methods to bound doses from thorium have been 

determined. Therefore, NIOSH has determined that unmonitored thorium doses at RFP cannot be 

reconstructed.

5.2.5 Neptunium

Neptunium processing at the RFP included preparation of pure neptunium oxide, metal and metal 

alloys, and the recovery of 237Np from a variety of residues (Conner and Baaso 1981). Processes 

included dissolution, anion exchange, precipitation, filtration, calcination, conversion to fluoride, and 

reduction to metal. Fabrication steps such as casting and rolling were also sometimes performed for 

the production of high-purity metal shapes and foils. Neptunium was recovered from residual 

materials including sand, slag, crucibles, casting skulls, and various alloys (with plutonium, tin, 

uranium, and zirconium).

Because of data issues and limitations, no specific methods to bound doses from neptunium have 

been determined. Therefore, unmonitored neptunium doses at the RFP cannot be reconstructed.

Attachment E provides additional details on neptunium operations.
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5.2.6 Critical Mass Laboratory

Building 886 housed the Critical Mass Laboratory (CML) at RFP. Mixed fission and activation 

products (MFAPs) in both the fuel and containment materials present an internal dose potential for 

personnel who might ingest or inhale them. CML staff submitted routine bioassay (urinalysis and 

whole-body counts) to detect intakes of plutonium, uranium, and/or americium, but MFAPs were not 

routinely monitored.

Attachment H provides a detailed analysis of the CML. Based on this analysis, no significant 

unmonitored exposure is associated from the generation of fission or activation products.

5.3 IN VITRO

5.3.1 Plutonium Urinalysis

5.3.1.1 Methods, Units, Isotopes, and Interferences

Through 1989, the units of the results are dpm/24-hr excretion period (dpm/24-hr sample). After 

1989, the units of the results are dpm/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 followup chelation using diethylene triamine pentaacetic acid 

(DTPA). Assume a 24-hour excretion period unless the record indicates that the actual excretion 

period was different.

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 from the air proportional detector system included activity from 238Pu, 239Pu, and 240Pu. 

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

Intake assessments are simpler and more favorable to claimants if dose reconstructors assume 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. This factor is 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.

Interferences were probably 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 probably included 

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

plutonium through 1973, the result could have included 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.

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. 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.
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DTPA and ethylene diamine tetraacetic acid (EDTA) chelation treatments cause enhanced excretion 

of plutonium in the urine. Urine data from within 90 days of a chelation injection have historically been 

excluded from calculations of intakes or depositions of plutonium. Information in the medical or 

dosimetry records should allow dose reconstructors 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 that was affected by chelation were flagged with a code 1. Code 1 was 

also used to flag urine data that did not pass quality standards. Dose reconstructors should be wary 

of any urine result flagged with a code 1 and in general should not use these data in dose 

reconstruction (Various 1965–1987, p. 28). 

5.3.1.2 Plutonium Reporting Levels, Minimum Detectable Amounts, 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 computergenerated 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 (KHC 2002).

The 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 dose reconstructors are not likely to have sufficient data to determine the sample-specific MDA, 

so the median values should be used.

Table 5-6 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 (Author 

unknown 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 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 that was contractually required in 

the Rocky Flats Environmental Technology Site bioassay statement of work (KHC 1998b) for any 

laboratory that processed the sample. 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 in Table 5-6 should be used.

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

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

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

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

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

apply to each sample, the MDAs in Table 5-6 do not account for the improvement until the transition 

was completed for all samples (i.e., the MDAs are favorable to claimants).
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Table 5-6. 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 rereported 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

based on an examination of the Comprehensive Epidemiology Data Resource (CEDR) database. 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).

5.3.2 Americium Urinalysis

5.3.2.1 Methods, Units, Isotopes, 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 dpm/24-hr excretion period through 1989. After 1989, the units of the 

results are dpm/sample regardless of the sample volume or excretion period. This statement is based 

on the observation that the same reporting format used for plutonium results was used for americium 

results.

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. This biases the result high and is considered favorable to 

the claimant. 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.

5.3.2.2 Americium Reporting Levels, Minimum Detectable Amounts, 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. Until it is determined, dose reconstructors should assume that the 

reporting level for 1968 to 1971 was continued through 1976, which is considered favorable to the 

claimant. 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-hour sample starting in about 1990. Instead, the results are 

dpm/sample, regardless of the sample volume (ORAUT 2003). 
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The MDAs for americium (Table 5-7) were determined as described for plutonium (see Section 5.3.1.2

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

Table 5-7. Median MDA values for americium.

a,b

Period dpm/24-hr sample

1963 0.44

1964–1965c 0.55

1965–1970c 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.

c. In overlapping years the more favorable MDAs should be assumed.

The discussions of MDA and uncertainty for plutonium urinalysis in Section 5.3.1.2 apply to americium 

urinalysis.

5.3.3 Uranium Urinalysis

5.3.3.1 Enriched Uranium

5.3.3.1.1 Methods, Units, Isotopes, and Interferences

The units of the results are dpm/24-hr excretion period for the entire period. Because urine samples 

analyzed for EU were counted with the air proportional detectors, all of the alpha-emitting isotopes of 

uranium are included in the result. Site-specific information about possible interferences that might 

have occurred for the urinalysis methods for EU is not available.

5.3.3.1.2 Reporting Levels, Minimum Detectable Amounts, and Uncertainties

Table 5-8 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).

Table 5-8. Median MDAs for EU. 

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 5.3.1.2 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).
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5.3.3.2 Depleted Uranium

5.3.3.2.1 Methods, Units, Isotopes, and Interferences

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

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

extraction procedure, and the electrodeposited extract 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 were processed at an offsite commercial laboratory according to provisions of the 

bioassay statement of work (KHC 1998b).

The units for 1952 to April 1964 were micrograms of uranium per 24-hour excretion period. The mass 

measurement was for all the isotopes of uranium. From May 1964 to 1989, the units were dpm/24-hr 

sample. After 1989, the units of the results were dpm/sample, regardless of the sample volume or 

excretion period (ORAUT 2003). 

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

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

measurements. For the 1980s, 238U contributes 89% of the alpha activity. Therefore, the logs have 

not been reviewed to determine the other uranium isotopes. In the 1990s, the urine data reports 

include the results separately for 234U, 235U, and 238U.

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

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

urine results for this background, which was highly variable.

5.3.3.2.2 Depleted Uranium Reporting Levels, Minimum Detectable Amounts, 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. 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 5-9. For alpha-counting methods, 

the MDAs in the period from April 1964 to 1971 are the same as those for EU in Table 5-7. The MDA 

value for 1972 to 1979 was extrapolated from the value for the previous period. The MDAs for 1980 

to the present were derived in the same manner as that for plutonium but are based on 238U. 

Table 5-10 lists median MDAs for DU from May 1964 to the present.

Table 5-9. Median MDAs for DU from 1952 to April 1964. 

Period μg/24-hr sample

1952–1955a 31

1955–1959a 12

1960–04/1964 11

a. In overlapping years the more favorable MDAs should be assumed.

The discussion of the uncertainty for plutonium in Section 5.3.1.2 applies to DU.
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Table 5-10. Median MDAs for DU from May 1964 to the 

present.

b

Period dpm/24-hr sample

05/1964–1969 31

1970–1971 25

1972–1979 25c

1980–1989 0.56

1990–1992 0.4a

1993–present 0.1a

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.

5.3.4 Gross Alpha Urinalysis

5.3.4.1 Methods, Units, Isotopes, and Interferences 

Gross alpha measurement is a nonspecific analysis that was 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, probably in 1973 based on an examination of the 

CEDR database. However, isolated gross alpha measurements were identified as late at January 

1980. 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. The analyzed isotopes were all of the alpha-emitting isotopes of the analyte. 

5.3.4.2 Reporting Levels, Minimum Detectable Amounts, 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 and credited to 

plutonium. Gross alpha data are coded as G in the urine data reports. 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). 

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

system to determine whether to credit the result to EU, to plutonium, or to 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 is considered an upper bound because of the

nonspecificity of the analysis.

The MDAs for gross alpha in Table 5-11 were determined as described in Attachment A.

Table 5-11. 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
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5.3.5 Unmonitored Tritium Exposure

The unmonitored tritium exposure rates in Table 5-12 should be assigned to all workers at RFP. 

Attachment G provides the basis for the exposure rates.

Table 5-12. Unmonitored tritium exposure rates. 

Period Dose rate (rem/yr)

Before 1973 0.0375

1973 0.103

After 1973 0.0

5.3.5.1 Reporting Levels, Minimum Detectable Amounts, and Uncertainties

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

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

measurements. It is assumed that liquid scintillation was used in the 1970s. 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 not been quantified for 

the methods in the 1970s and 1980s, it is probably in the range of several hundred to several 

thousand picocuries per liter (AEC 1973, ca. 1974; Various 1955–1976; RFP 1952–1971). The MDA 

for tritium should be assumed to be 600 pCi/L for all tritium bioassay (KHC 1998c, p. 176).

5.3.6 Alvin E. Blackwell Bioassay Laboratory

In late 1991, RFP identified an issue with their outside bioassay laboratory Alvin E. Blackwell 

Bioassay Laboratory (AEB). The RFP Internal Dosimetry Committee meeting notes indicate that the 

site determined that they needed to send a set of quality control samples to AEB before they send a 

group of baseline samples for analysis to determine if AEB is qualified. Six of the eight quality control 

samples sent to AEB failed the data quality objectives (EG&G 1992).

In a review of individual dosimetry records, sample results assessed by AEB included a cover letter 

that states that these results are not to be considered valid data. In addition, all personnel would be 

resampled. 

Any bioassay results analyzed by AEB should not be considered valid. The individual should be 

considered unmonitored during this period unless non-AEB bioassay results are available. 

5.4 IN VIVO

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 that are 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 low-energy X-rays (L X-rays). Rather, 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).
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The RFP lung counter also measured 234Th, using 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 under the assumption of equilibrium (Berger 1988a).

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

Section 5.6 discusses the data, and Attachment C contains examples of the report forms.

5.4.1 Americium and Plutonium

5.4.1.1 Methods, Units, Isotopes, and Interferences

Before May 1995, 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 and accompanied by a decision that is 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. Figure C-18 is an example of the implementation of these 

modifications. In addition, the activity of 241Am was stated as a fraction of the MPLB, which was 14.7 

nCi (Falk ca. 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. The second interference is the contribution of count from 241Am not in the lungs (e.g., 

contributions from contamination on the skin, activity in the lymph nodes, etc.).

5.4.1.2 Reporting Levels, Minimum Detectable Amounts, and Uncertainties

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 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. 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. The decision level for 

1995 and later was calculated by ABACOS-Plus for a probability of a Type I (false positive) error of 

5% (KHC 2000b, p. 90). The decision level was used as a reporting level from May 1995 to midFebruary 1997. For ABACOS-Plus lung count reports, from May 1995 to mid-February 1997, when a 

result above the decision level is reported, the MDA is assumed to be twice the decision level. The 

decision level is determined as 1.645 times the 1-sigma standard deviation of the measurement. Note 

that during this time the 2-sigma percent error was normally reported. Therefore the MDA would 

equal to 1.645 × (measured activity) × (2-sigma percent error) ÷ (100).

Table 5-13 lists the MDAs for 241Am, which were calculated for the evolution of lung-counting systems 

at RFP as described in Attachment B.
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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, dose 

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

extrapolate) from the values in Table 5-13. 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 mid-February 1997 and later. (The MDA values are reported on report forms from midFebruary 1997 and later, but the values are not worker-specific. Dose reconstructors should 

disregard these MDA values.) The default MDA would be for an index of 1.35 (median value for a 

range of 0.9 to 1.8) if height and weight (or index) data for the worker are not available. Dose 

reconstructors should assume the MDA is twice the decision level for mid-February 1997 and later 

lung count reports that include the non-worker-specific MDA.

The MDA for plutonium should 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 

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

The uncertainties of the results were reported for the net counts 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. Figures C-20 and C-21 are examples of the implementation of these modifications. With the 

advent of ABACOS-Plus in 1995, the percent error at 2 standard deviations was reported for all 

identified nuclides until around mid-February 1997, and the percent error at 1 standard deviation was 

reported for all identified nuclides after that time. 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 (KHC 2000b, p. 89).

5.4.2 Thorium and Depleted Uranium

5.4.2.1 Methods, Units, Isotopes, 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 mid-February 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 
 

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Table 5-13. Summary of MDAs (nCi) for 241Am.

Perioda Detector system Index

Minimum system 

half time

Minimum system 

full time

Standard system

half time

Standard system 

full time

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

1964–1968 NaI(Tl) 4 × 4 1.35 2.8 2.5 2.1 1.9

1964–1968 NaI(Tl) 4 × 4 1.80 4.6 4.1 3.5 3.2

1969→ NaI(Tl) 4 × 4 0.90 Not applicable Not applicable 0.8 0.76

1964–1968 NaI(Tl) 4 × 4 1.35 Not applicable Not applicable 1.3 1.3

1964–1968 NaI(Tl) 4 × 4 1.80 Not applicable Not applicable 2.2 2 

1973→ Phoswich 0.90 Not applicable Not applicable 1.2 1.2

1973→ Phoswich 1.35 Not applicable Not applicable 2.0 2.

1973→ Phoswich 1.80 Not applicable Not applicable 3.3 3.2

1976–1978 Ortec Arrays (high-purity Ge) 0.90 0.26 0.18 0.2 0.14

1976–1978 Ortec Arrays (high-purity Ge) 1.35 0.48 0.32 0.37 0.25

1976–1978 Ortec Arrays (high-purity Ge) 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

1979→ Ortec Arrays (high-purity Ge) 1.35 0.37 0.25 0.28 0.19

1979→ Ortec Arrays (high-purity Ge) 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

1978→ PGT I Arrays (high-purity Ge) 1.35 0.4 0.27 0.31 0.21

1978→ PGT I Arrays (high-purity Ge) 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

1979→ PGT I Arrays (high-purity Ge) 1.35 0.31 0.21 0.24 0.16

1979→ PGT I Arrays (high-purity Ge) 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

1979→ PGT II Arrays (high-purity Ge) 1.35 0.4 0.28 0.31 0.21

1979→ PGT II Arrays (high-purity Ge) 1.80 0.74 0.5 0.57 0.39

1985→ PGT Organ Pipe Ge Detectors 0.90 Not applicable Not applicable 0.15 0.11

1985→ PGT Organ Pipe Ge Detectors 1.35 Not applicable Not applicable 0.26 0.18

1985→ PGT Organ Pipe Ge Detectors 1.80 Not applicable Not applicable 0.46 0.32

1991→ EG&G Organ Pipe Ge Detectors 0.90 Not applicable Not applicable 0.14 0.1

1985→ PGT Organ Pipe Ge Detectors 1.35 Not applicable Not applicable 0.26 0.18

1985→ PGT Organ Pipe Ge Detectors 1.80 Not applicable Not applicable 0.48 0.33

1995→ Ortec 2 Organ Pipe Ge Detectors 0.90 Not applicable Not applicable Not applicable 0.14

1985→ PGT Organ Pipe Ge Detectors 1.35 Not applicable Not applicable Not applicable 0.3

1985→ PGT Organ Pipe Ge Detectors 1.80 Not applicable Not applicable Not applicable 0.6

of 263

a. In overlapping years, the more favorable MDAs should be assumed.
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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 5.2.3.2.1).

5.4.2.2 Reporting Levels, Minimum Detectable Amounts, and Uncertainties

Reporting levels were not generally used for DU until 1995 with the implementation of ABACOS-Plus 

(see Section 5.4.2.1). Before 1995, the 238U activity was generally quantified only after verification of 

an intake. For ABACOS-Plus lung count reports, from May 1995 to mid-February 1997, when a result 

above the decision level is reported, the MDA is assumed to be twice the decision level. The decision 

level is determined as 1.645 times the 1-sigma standard deviation of the measurement. Note that 

during this time the 2-sigma percent error was normally reported. Therefore the MDA would equal to 

1.645 × (measured activity) × (2-sigma percent error) ÷ (100).

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 5.4.2.1 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. Dividing the 238U MDA by 0.89 accounts for the contribution to the 

DU MDA from activities of the other uranium isotopes.

As noted in Section 5.4.1.2 for americium and plutonium, MDA values are reported on forms for midFebruary 1997 and later, but are not worker-specific. Dose reconstructors should disregard these 

MDA values. Dose reconstructors should assume the MDA is twice the decision level for midFebruary 1997 and later lung count reports that include the non-worker-specific MDA.

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. If a superequilibrium 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 approach is favorable to claimants. 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 days after a thorium strike, and 90% occurs after 

80 days.

5.5 OTHER BIOASSAY DATA

5.5.1 Wound Count 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 RFP 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 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.
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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.

Guidance on assessing wound intakes is provided in ORAUT-OTIB-0022, Guidance on Wound 

Modeling for Internal Dose Reconstruction (ORAUT 2005a). 

5.5.2 Nasal Smears and Fecal Samples

Nasal smear (later called swab) and fecal sample data were occasionally performed 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 were also 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 after 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 (KHC 

1998d, p. 62).

The reported MDAs (KHC 1998d, pp. 67–68) are:

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

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

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

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

• 100 dpm/sample for fecal samples with a 2-day 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, which is approximately one-half of the sample-specific MDA. MDA values for 

earlier years are not available.

5.6 RECORDS AND REPORTS

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

bioassay data. Attachment C, Examples of Records and Reports Used at RFP, contains the figures 

described below.
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5.6.1 Urinalysis Records and Reports

Figures C-1 to C-3 are examples of the Urinalysis Record Card and the HSDS – 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.

5.6.2 Interpretation of the Urinalysis Record Card

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

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

either a number or BK (see Section 5.3.1.2). 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 HSDS – 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 5.3.1.2 (i.e., some plutonium results reported as BK on the 

card were rereported with the actual result). The urine data record written on the Urinalysis Record 

Card preceded the HSDS and was the probable 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 be considered the correct data, as recommended. 

On some cards, dose reconstructors might observe the initially reported result was 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. 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.

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).

5.6.3 Interpretation of the Health Sciences Data System – Urinalysis Detail Report

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

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

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

ANAL is the code for the analyte:

P = plutonium,

A = americium,

U = EU (pre-1970, approximately),
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U = DU (1970–1989, approximately),

D = DU (1952–1969, approximately), and

G = gross alpha.

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

0 = use normally in the calculation;

1 = do not use in the calculation; and

2 = date of a new intake.

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

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

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

90 days of a chelation treatment were generally (or should have been) coded as 1 (Various 1965–

1987, p. 28). The use of Code 2 to flag the date of a new significant intake occurred inconsistently. In 

reports from the 1980s, an asterisk was used instead of a Code 2 to flag the date of a new intake. 

Dose reconstructors 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. This circumstance 

is evident in the example reports in Figures C-2, C-3, and especially C-5.

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

systemic burden that was calculated from Code 0 results for plutonium and for americium. 

This data field is not likely to be of use.

5.6.4 Interpretation of Other Urinalysis 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 for plutonium analysis, and Figure C-7 is 

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

columns are self-explanatory; the remaining columns are:

• Dec Level is the decision level 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 previous Code 0 for urine data in 

the HSDS. Codes 1, 3, and 4 indicate a failed analysis and disqualify the result (Various 

1982–2005, p. 160). 

• DQO, for “data quality objective,” 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. Therefore, the ANN 

notation 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 (Various 1982–
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2005, p. 160). If the batch failed, every sample in the batch was conditionally failed pending 

further evaluation.

• 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.

• Analyte is self-explanatory.

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

• DPM is the result of the sample in dpm/sample. Dose reconstructors should assume a 

24-hour urine sample unless there is information that indicates otherwise. 

• 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. 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 dose reconstructors is to use the date written on 

the report in this case. The result header is largely self-explanatory.

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

dpm/sample (REPORT UNIT).

• The decision level 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.
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• 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 5.3.3.1.1, 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 is V). Use only results with a QUAL of V.

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

information is largely self-explanatory.

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

0600 was used as a default end time of the 24-hour excretion period if the actual end time was 

not documented.

• The VF is the volume fraction, the fraction of the sample that was analyzed. A volume fraction 

of 1 indicates that the entire sample was analyzed.

• Use only data that have a Data Validation Code of V.

Figure C-13 is an example of the data card that was 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 based on an examination of the CEDR database. 

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.

5.6.5 Lung Count Records and Reports

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.

• 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 probably use the term “body 

count” instead of “lung count.” Dose reconstructors 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 as 40 minutes live time (MLT).

• 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 plutonium alpha emitters). 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/wk), organ mass (m) = 

1,000 g, f2 = 1, and ε = 53 (ICRP 1959, Table 5; and based on a relative biological 

effectiveness = 10).
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• The Body Location is the position of the detector. In this case, the detectors were over the 

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

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

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.

• 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. Dose reconstructors should 

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

lung dataset generally includes contributions from both right and left lungs.

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

unknown case (see Section 5.4.1.2). 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 dose reconstructors 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 is 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.

• 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 that was 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.

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 only on 

counting statistics. For workers with confirmed lung depositions, the calibration factors for plutonium 

and americium were generally written on the form, as in this case.
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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 americium, or the lung burden. 

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

the cutoff (see also Figure C-22 for example).

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. 

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 probably 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 (see also Figure C-26 for example). ROI 5 (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, namely low-end electronic noise in one or more of the 

detectors. Dose reconstructors should disregard all L X-ray data (including the 13- and 17-keV ROIs). 

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.

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. 

Figure C-29 is an example of the lung count report from an early version of ABACOS-Plus that was 

used through 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. This statement is the result of direct observation of 

information in Figure C-29. For ABACOS-Plus lung count reports, from May 1995 to mid-February

1997, when a result above the decision level is reported, the MDA is assumed to be twice the decision 

level. The decision level is determined as 1.645 times the 1-sigma standard deviation of the 

measurement. Note that during this time the 2-sigma percent error was normally reported. Therefore 

the MDA would equal to 1.645 × (measured activity) × (2-sigma percent error) ÷ (100).

Figure C-30 is an example of the lung count report from ABACOS-Plus 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 
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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 ABACOS-Plus for a worker with a confirmed 

deposition. The software calculated the deposition for the plutonium isotopes based on the intake 

date 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 was assigned to the 

plutonium isotopes. The value of the ppm 241Am on the date of the count was not reported on lung 

count reports that were generated by ABACOS-Plus. This value can be calculated using Equation B18 in Attachment B.

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

Analysis Limits, and the total activity.

Dose reconstructors 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.

Dose reconstructors should be aware that the lung counter detectors were also used for wound 

counts (Berger 1988b; KHC 2000b, p. 93). 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 calculated activities for plutonium for lung counts were 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 dose reconstructors choose to use another intake date or 

initial ppm 241Am, they 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. Dose reconstructors should adjust the activities 

for the discontinuity factors presented in Attachment B. In general, use of the discontinuity factors is 

favorable to claimants. The only exception to this statement 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.

5.7 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
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FBI (Federal Bureau of Investigation), 1989i, Interview with Rocky Flats Employee Joseph Krupar, 

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FBI (Federal Bureau of Investigation), 1991a, Interview with Former Rocky Flats Employee Roger 

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FBI (Federal Bureau of Investigation), 1991b, Interview with Department of Energy Employee Lawton 

J. O’Neil, U.S. Department of Justice, Denver, Colorado, March 27. [SRDB Ref ID: 141293]

FBI (Federal Bureau of Investigation), 1991c, Interview with Rocky Flats Employee Joseph Krupar, 

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FBI (Federal Bureau of Investigation), 2001, Federal Bureau of Investigation Raid, 1989 Desert Glow 

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Hammond, S. E, 1956–1958, collection of Monthly Progress Reports, Chemistry Lab and Personnel 

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Hammond, S. E., 1958, collection of Monthly Progress Reports, Chemistry Lab and Personnel Meters 

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Hammond, S. E., 1966b, “Status Report -- Industrial Hygiene & Bio-Assay -- July, 1966,” 

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Uranium, ANSI/HPS N13.22-1995, American National Standards Institute, McLean, Virginia, 

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HPS (Health Physics Society), 1996, An American National Standard – Performance Criteria for 

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Alamos, New Mexico. [SRDB Ref ID: 953]

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Uranium, Version 3, Rocky Flats Environmental Technology Site, Golden, Colorado, April 1. 

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KHC (Kaiser-Hill Company), 1998b, Statement of Work for Analytical Measurements, General 

Bioassay Services, Module BA01-B.4, Rocky Flats Environmental Technology Site, Golden, 

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KHC (Kaiser-Hill Company), 1998c, Internal Dosimetry Technical Basis Documentation, Section 7, 

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KHC (Kaiser-Hill Company), 1998d, Internal Dosimetry Technical Basis Documentation, In-Vitro 

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Energy’s Rocky Flats Plant, 1953–1993, Rocky Flats Environmental Technology Site, Golden, 

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KHC (Kaiser-Hill Company), 2000b, Internal Dosimetry Technical Basis Documentation, Section 3, InVivo Radiobioassay, Version 4, Rocky Flats Environmental Technology Site, Golden, 

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KHC (Kaiser-Hill Company), 2002, Internal Dosimetry Technical Basis Documentation, Section 5, 

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Kusmierczak, J. J., 2007, untitled letter to L. Elliot (National Institute for Occupational Safety and 

Health), Simmons Cooper, East Alton, Illinois, July 24. [SRDB Ref ID: 39868]

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Lawrence, J. N. P, 1984, Uranium Internal Exposure Evaluation Based on Urine Assay Data, LA10246-MS, University of California, Los Alamos National Laboratory, Los Alamos, New 

Mexico, September. [SRDB Ref ID: 7971]

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Radioactivity”, memorandum to G. Ansel, The Dow Chemical Company, Madison, Illinois, April 

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The Dow Chemical Company, Rocky Flats Plant, Golden, Colorado, February 7. [SRDB Ref 

ID: 24692, p. 59]

Mann, J. R., 1969, “Status Report -- Dosimetry -- April, 1969,” memorandum to C. W. Piltingarud, The 

Dow Chemical Company, Rocky Flats Plant, Golden, Colorado, May 7. [SRDB Ref ID: 

26463, p. 9]

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. [SRDB Ref ID: 

7974]

Martella, L. L., R. H. Guyer, and W. C. Leak, 1987, Production-Scale Plutonium-Neptunium 

Separation and Residue Recovery at Rocky Flats Plant, RFP-4030, R. L. Thomas, editor, 

Rockwell International, Rocky Flats Plant, Golden, Colorado, May 26. [SRDB Ref ID: 129512]

Meyers, L., 2013, “RFP MT82 Inventories Spreadsheet,” spreadsheet, Protection Strategies 

Incorporated. [SRDB Ref ID: 126758]

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[SRDB Ref ID: 127264]

Mitchell, C. A., 1957, “Dust Samples Taken During Hand Sanding of HK31 Alloy,” memorandum to L. 

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40017]

Mitchell, C. A., 1958a, “Air Samples Taken During HK31 Sanding,” memorandum, The Dow Chemical 

Company, Madison, Illinois, March 17. [SRDB Ref ID: 39980, p.2]

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Sanding Discs,” memorandum to L. Silverstein, The Dow Chemical Company, Madison, 

Illinois, March 24. [SRDB Ref ID: 51843]

Moment, R. L, F. E. Gibbs, and C. J. Freiboth, 1999, History of Uranium-233 (233U) Processing at the 

Rocky Flats Plant, RS-090-056, ICF Kaiser Engineers, Los Alamos, New Mexico, April 1. 

[SRDB Ref ID: 22101]

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

Implementation Guideline, OCAS-IG-002, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August 21. [SRDB Ref ID: 22402]
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NIOSH (National Institute for Occupational Safety and Health), 2006, SEC Petition Evaluation Report, 

Petition SEC-00030, Rev. 0, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

April 7. [SRDB Ref ID: 132776]

NIOSH (National Institute for Occupational Safety and Health), 2010, Radiation Exposures Covered 

for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, DCAS-IG-003, Rev. 1, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, October 5. [SRDB Ref ID: 88929]

NIOSH (National Institute for Occupational Safety and Health), 2011, Site Profiles for Atomic 

Weapons Employers That Worked Uranium Metals, Battelle-TBD-6000, Rev. 1, Office of 

Compensation Analysis and Support, Cincinnati, Ohio, June 17. [SRDB Ref ID: 101251]

NIOSH (National Institute for Occupational Safety and Health), 2013a, SEC Petition Evaluation 

Report, Petition SEC-00192, Rev. 1, Office of Compensation Analysis and Support, Cincinnati, 

Ohio, September 30. [SRDB Ref ID: 132777]

NIOSH (National Institute for Occupational Safety and Health), 2013b, “Interview, [Redacted], 

12/13/2013,” documented communication with former Rocky Flats Plant worker, Cincinnati, 

Ohio, December 13. [SRDB Ref ID: 129782]

NIOSH (National Institute for Occupational Safety and Health), ca. 2013, untitled documented 

communication with former Rocky Flats Plant worker, Cincinnati, Ohio. [SRDB Ref ID: 

130248]

Norton, M., 1989, “Criminal Investigation Concerning the Rocky Flats Plant,” letter to W. H. Moor (U.S. 

Department of Energy, Deputy Secretary), U.S. Department of Justice, District of Colorado, 

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OMB (Office of Management and Budget), 2012, Fiscal Year 2012, Historical Tables, Budget of the 

U.S. Government, Executive Office of the President of the United States, Washington, D.C. 

[SRDB Ref ID: 129484]

ORAUT (Oak Ridge Associated Universities Team), 2003, Minimum Detectable Activity for Urinalysis 

Methods at Rocky Flats, Oak Ridge, Tennessee, June 5. [SRDB Ref ID: 23634]

ORAUT (Oak Ridge Associated Universities Team), 2004, Coworker Data Exposure Profile 

Development, ORAUT-PLAN-0014, Rev. 00, Oak Ridge, Tennessee, November 24. [SRDB 

Ref ID: 166559]

ORAUT (Oak Ridge Associated Universities Team), 2005a, Guidance on Wound Modeling for Internal 

Dose Reconstruction, ORAUT-OTIB-0022, Rev. 00, Oak Ridge, Tennessee, November 18. 

[SRDB Ref ID: 19443]

ORAUT (Oak Ridge Associated Universities Team), 2005b, Analysis of Coworker Bioassay Data for 

Internal Dose Assignment, ORAUT-OTIB-0019, Rev. 01, Oak Ridge, Tennessee, October 7. 

[SRDB Ref ID: 19438]

ORAUT (Oak Ridge Associated Universities Team), 2006, Estimating the Maximum Plausible Dose to 

Workers at Atomic Weapons Employer Facilities, ORAUT-OTIB-0004, Rev. 03 PC-02, Oak 

Ridge, Tennessee, December 6. [SRDB Ref ID: 36191]
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ORAUT (Oak Ridge Associated Universities Team), 2007, Rocky Flats Plant – Site Description, 

ORAUT-TKBS-0011-5, Rev. 01, Oak Ridge, Tennessee, February 1. [SRDB Ref ID: 30013]

ORAUT (Oak Ridge Associated Universities Team), 2013a, “The White Paper on Np Processing...,” 

documented communication with U.S. Department of Energy employee, Oak Ridge, 

Tennessee, December 13. [SRDB Ref ID: 129510]

ORAUT (Oak Ridge Associated Universities Team), 2013b, Data Collection Document Discover and 

Review, Oak Ridge, Tennessee, December 13. [SRDB Ref ID: 129610]

ORAUT (Oak Ridge Associated Universities Team), 2013c, “Materials and Construction of SST and 

ATMX Transports (Interview Conducted at EMCBC Denver,” documented communication with 

U.S. Department of Energy employee, Oak Ridge, Tennessee, December 12. [SRDB Ref ID: 

129509]

ORAUT (Oak Ridge Associated Universities Team), 2013d, Data Collection Document Discover and 

Review, Oak Ridge, Tennessee, December 12. [SRDB Ref ID: 129608]

ORAUT (Oak Ridge Associated Universities Team), 2013e, Data Collection Document Discover and 

Review, Oak Ridge, Tennessee, December 12. [SRDB Ref ID: 129613]

ORAUT (Oak Ridge Associated Universities Team), 2013f, “Secure Transport Design and Fabrication 

at Rocky Flats,” documented communication with former Rocky Flats Plant worker, Oak Ridge, 

Tennessee, December 13. [SRDB Ref ID: 129511]

ORAUT (Oak Ridge Associated Universities Team), 2013g, “Documented Communication with 

[Redacted] on Rocky Flats Plant, December 13, 2013,” documented communication with 

former Rocky Flats Plant worker, Oak Ridge, Tennessee, December 13. [SRDB Ref ID: 

130557]

ORAUT (Oak Ridge Associated Universities Team), 2013h, “Secure Transport Design and Fabrication 

at Rocky Flats,” documented communication with former Rocky Flats Plant worker, Oak Ridge, 

Tennessee, December 13. [SRDB Ref ID: 129511]

ORAUT (Oak Ridge Associated Universities Team), 2013i, “Documented Communication SEC-00192 

with [Redacted] on the FBI Investigation at Rocky Flats, March 14, 2013,” documented 

communication with former Rocky Flats Plant worker, Oak Ridge, Tennessee, March 14. 

[SRDB Ref ID: 123338]

ORAUT (Oak Ridge Associated Universities Team), 2013j, “Documented Communication with 

[Redacted] on the Rocky Flats Plant July 18, 2013 - SEC000192,” documented communication 

with former Rocky Flats Plant worker, Oak Ridge, Tennessee, July 18. [SRDB Ref ID: 

126995]

ORAUT (Oak Ridge Associated Universities Team), 2013k, “Documented Communication with 

[Redacted] on Rocky Flats Plant August 20, 2013, SEC-00192,” documented communication 

with former Rocky Flats Plant worker, Oak Ridge, Tennessee, August 20. [SRDB Ref ID: 

127272]

ORAUT (Oak Ridge Associated Universities Team), 2013l, “Documented Communication with 

[Redacted] on Rocky Flats Plant, November 26, 2013,” documented communication with 

former Rocky Flats Plant worker, Oak Ridge, Tennessee, November 26. [SRDB Ref ID: 

129519]
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ORAUT (Oak Ridge Associated Universities Team), 2013m, “Interview on the FBI Raid at RFP,” 

documented communication with former Rocky Flats Plant worker, Oak Ridge, Tennessee, 

December 10. [SRDB Ref ID: 131488]

ORAUT (Oak Ridge Associated Universities Team), 2014a, “Documented Communication with 

[Redacted] on Rocky Flats Plant, February 6, 2014,” documented communication with former 

Rocky Flats Plant worker, Oak Ridge, Tennessee, February 6. [SRDB Ref ID: 130877]

ORAUT (Oak Ridge Associated Universities Team), 2014b, “Re-interview with [Redacted] on 

Neptunium at Rocky Flats Plant,” documented communication with former Rocky Flats Plant 

worker, Oak Ridge, Tennessee, December 12. [SRDB Ref ID: 138682]

ORAUT (Oak Ridge Associated Universities Team), 2014c, “Documented Communication with 

[Redacted] on Rocky Flats Plant, February 12, 2014,” documented communication with former 

Rocky Flats Plant worker, Oak Ridge, Tennessee, February 12. [SRDB Ref ID: 131225]

ORAUT (Oak Ridge Associated Universities Team), 2014d, “Interview on Neptunium with [Redacted],” 

documented communication with former Rocky Flats Plant workers, Oak Ridge, Tennessee, 

December 12. [SRDB Ref ID: 138666]

ORAUT (Oak Ridge Associated Universities Team), 2014e, “SEC-00192 Rocky Flats Neptunium,” 

spreadsheet, Oak Ridge, Tennessee, September. [SRDB Ref ID: 137075]

ORAUT (Oak Ridge Associated Universities Team), 2014f, “Documented Communication with 

[Redacted] on Rocky Flats Plant, February 12, 2014,” documented communication with former 

Rocky Flats Plant worker, Oak Ridge, Tennessee, February 12. [SRDB Ref ID: 131005]

ORAUT (Oak Ridge Associated Universities Team), 2014g, “Interview with [Redacted] about 1989 FBI 

Rocky Flats Raid,” documented communication with former Rocky Flats Plant worker, Oak 

Ridge, Tennessee, May 14. [SRDB Ref ID: 133781]

ORAUT (Oak Ridge Associated Universities Team), 2014h, “Follow-up Call for Documents Submitted 

for Rocky Flats Plant,” documented communication with Rocky Flats Plant worker, Oak Ridge, 

Tennessee, June 4. [SRDB Ref ID: 132787]

ORAUT (Oak Ridge Associated Universities Team), 2014i, “Documented Communication with 

[Redacted] on Rocky Flats Plant, February 13, 2014,” documented communication with former 

Rocky Flats Plant worker, Oak Ridge, Tennessee, February 13. [SRDB Ref ID: 130943]

ORAUT (Oak Ridge Associated Universities Team), 2014j, “Documented Communication with 

[Redacted] on Rocky Flats Plant, January 21, 2014,” documented communication with former 

Rocky Flats Plant worker, Oak Ridge, Tennessee, January 21. [SRDB Ref ID: 131006]

ORAUT (Oak Ridge Associated Universities Team), 2014k, “Interview on the FBI Raid at RFP,” 

documented communication with former Rocky Flats Plant worker, Oak Ridge, Tennessee, 

May 13. [SRDB Ref ID: 133401]

ORAUT (Oak Ridge Associated Universities Team), 2014l, “Interview on the FBI Raid at RFP,” 

documented communication with former Rocky Flats Plant worker, Oak Ridge, Tennessee, 

May 21. [SRDB Ref ID: 132826]
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ORAUT (Oak Ridge Associated Universities Team), 2014m, “Interview on the FBI Raid at RFP,” 

documented communication with former Rocky Flats Plant criticality panel member, Oak 

Ridge, Tennessee, May 22. [SRDB Ref ID: 132825]

ORAUT (Oak Ridge Associated Universities Team), 2015a, “Documented Communication with 

[Redacted] on the Critical Mass Laboratory at Rocky Flats Plant, October 22, 2015,” 

documented communication with former RFP worker, Oak Ridge, Tennessee, October 13. 

[SRDB Ref ID: 150602]

ORAUT (Oak Ridge Associated Universities Team), 2015b, Fission and Activation Product 

Assignment for Internal Dose-Related Gross Beta and Gross Gamma Analyses, ORAUTOTIB-0054, Rev. 04, Oak Ridge, Tennessee, August 27. [SRDB Ref ID: 137075]

ORAUT (Oak Ridge Associated Universities Team), 2015d, Technical Basis Document for the 

Hanford Site – Occupational Internal Dose, Rev 06, ORAUT-TKBS-0006-5, November 16. 

[SRDB Ref ID: 148938]

ORAUT (Oak Ridge Associated Universities Team), 2016, Use of Claimant Datasets for Coworker 

Modeling, Rev. 01, ORAUT-OTIB-0075, Oak Ridge, Tennessee, June 17. [SRDB Ref ID: 

157060]

ORAUT (Oak Ridge Associated Universities Team), 2018, Internal Dose Reconstruction, ORAUTOTIB-0060, Rev. 02, Oak Ridge, Tennessee, April 20. [SRDB Ref ID: 171554]

ORAUT (Oak Ridge Associated Universities Team), 2020, Estimating Doses for Plutonium Strongly 

Retained in the Lung, ORAUT-OTIB-0049, Rev. 02, Oak Ridge, Tennessee, September 1. 

[SRDB Ref ID: 178329] 

Owen, M. W., 2007, “Freedom of Information Act (FOIA) Request F2007-00179,” letter to J. 

Kusmierczak (Simmons Cooper), U.S. Department of Energy, Office of Legacy Management, 

Washington, D.C., July 13. [SRDB Ref ID: 34093]

Palmer, B., 2004, Minimally Reflected Cylinders of Highly Enriched Solutions of Uranyl Nitrate, 

NEA/NSC/DOC(95)03/II, Volume II, Idaho Chemical Processing Plant, Idaho Falls, Idaho, 

September 30. [SRDB Ref ID: 142464]

Peloubet, J. A.,1959, untitled letter to L. R. Rogers (U.S. Atomic Energy Commission, Radiation 

Safety Branch), The Dow Chemical Company, Rocky Flats Plant, Golden, Colorado, August 

28. [SRDB Ref ID: 29584]

RFP (Rocky Flats Plant), 1952–1971, collection of monthly reports, Golden, Colorado. [SRDB Ref ID: 

24686]

RFP (Rocky Flats Plant), 1965–1986, collection of Positive Plutonium Body Counter Results, Golden, 

Colorado. [SRDB Ref ID: 137387]

RFP (Rocky Flats Plant), 1965–2005, collection of Rocky Flats Body Counter Data, Golden, Colorado. 

[SRDB Ref ID: 138540] 

RFP (Rocky Flats Plant), 1975–2004, collection of interviews, correspondence and waste stream 

identification reports, Golden, Colorado. [SRDB Ref ID: 142722]
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RFP (Rocky Flats Plant), 1976–1996, collection of radiation monitoring procedures, Golden, Colorado. 

[SRDB Ref ID: 26563] 

RFP (Rocky Flats Plant), 1984–1990, collection of chain of custody forms. [SRDB Ref ID: 127117]

RFP (Rocky Flats Plant), 1994, Routine Bioassay Monitoring Program, 1-H04-HSP-18.20, Rev. 0, 

Golden, Colorado, March 18. [SRDB Ref ID: 24342]

Richardson, R. B., and D. W. Dunford, 2001, “Review of the ICRP Tritium and C-14 Internal 

Dosimetry Models and Their Implementation in the Genmod-PC Code,” Health Physics, 

volume 81, number 3, pp. 289–301. [SRDB Ref ID: 172475]

Rigsby, A. W. Jr., 1997, “United States of America v. Rockwell Case Number 89-M-1154,” letter to K. 

DuWaldt (Williams, Youle & Koenigs), U.S. Department of Energy, Rocky Flats Field Office, 

Golden, Colorado, September 22. [SRDB Ref ID: 127261]

RMRS (Rocky Mountain Remediation Services), 1999, Bioassay Sample Program Tracking and 

Trending, RF/RMRS-99-[incomplete], Rev. 0, Rocky Flats Environmental Technology Site, 

Golden, Colorado, August. [SRDB Ref ID: 24337]

Rockwell International, 1967, collection of experiments and calibration notes, Rocky Flats Plant, 

Golden, Colorado. [SRDB Ref ID: 157753]

Rockwell International, 1976, collection of experiment and calibration notes, Rocky Flats Plant, 

Golden, Colorado. [SRDB Ref ID: 157751]

Rockwell International, 1977a, Experiment at R.F.N.S.F., 1965 to April 19, 1977, Rocky Flats Plant, 

Golden Colorado. [SRDB Ref ID: 157757]

Rockwell International, 1977b, collection of experiment notes, Rocky Flats Plant, Golden, Colorado. 

[SRDB Ref ID: 157748]

Rockwell International, 1977c, Uranium Isotopic Analysis Report, Building 886, Rocky Flats Plant, 

Golden, Colorado, August 7. [SRDB Ref ID: 157754]

Rockwell International, 1980, collection of Routine Air Sample Results, Buildings 883 and 886, Rocky 

Flats Plant, Golden, Colorado, December. [SRDB Ref ID: 159122]

Rockwell International, 1982a, collection of Building 886 radiological surveys, Rocky Flats Plant, 

Golden, Colorado, September. [SRDB Ref ID: 152267]

Rockwell International, 1982b, collection of Building 886 radiological surveys, Rocky Flats Plant, 

Golden, Colorado, June . [SRDB Ref ID: 152295]

Rockwell International, 1983a, collection of Routine Air Sample Results, Buildings 865 and 886, 

Rocky Flats Plant, Golden, Colorado, May. [SRDB Ref ID: 159136]

Rockwell International, 1983b, collection of Routine Air Sample Results, Buildings 865 and 886, 

Rocky Flats Plant, Golden, Colorado, September. [SRDB Ref ID: 159154]

Rockwell International, 1983c, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, October. [SRDB Ref ID: 156834]
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Rockwell International, 1984a, collection of Building 886 radiological surveys, Rocky Flats Plant, 

Golden, Colorado, January and February. [SRDB Ref ID: 152277]

Rockwell International, 1984b, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, July. [SRDB Ref ID: 156857]

Rockwell International, 1984c, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, August. [SRDB Ref ID: 156858]

Rockwell International, 1984d, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, September. [SRDB Ref ID: 156859]

Rockwell International, 1984e, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, October. [SRDB Ref ID: 156860]

Rockwell International, 1984f, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, November. [SRDB Ref ID: 156861]

Rockwell International, 1984g, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, December. [SRDB Ref ID: 156862]

Rockwell International, 1985a, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, January. [SRDB Ref ID: 156863]

Rockwell International, 1985b, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, February. [SRDB Ref ID: 156867]

Rockwell International, 1985c, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, March. [SRDB Ref ID: 156876]

Rockwell International, 1985d, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, April. [SRDB Ref ID: 156878]

Rockwell International, 1985e, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, May. [SRDB Ref ID: 160323] 

Rockwell International, 1985f, collection of Routine Air Sample Results, Buildings 865 and 886, Rocky 

Flats Plant, Golden, Colorado, June. [SRDB Ref ID: 159142]

Rockwell International, 1985g, collection of Routine Air Sample Results, Buildings 865 and 886, 

Rocky Flats Plant, Golden, Colorado, July. [SRDB Ref ID: 159143]

Rockwell International, 1985h, collection of Routine Air Sample Results, Buildings 865 and 886, 

Rocky Flats Plant, Golden, Colorado, August. [SRDB Ref ID: 159144]

Rockwell International, 1985i, collection of Routine Air Sample Results, Buildings 865 and 886, Rocky 

Flats Plant, Golden, Colorado, September. [SRDB Ref ID: 159145]

Rockwell International, 1985j, collection of Routine Air Sample Results, Buildings 865 and 886, Rocky 

Flats Plant, Golden, Colorado, October. [SRDB Ref ID: 159146]

Rockwell International, 1985k, collection of Routine Air Sample Results, Buildings 865 and 886, 

Rocky Flats Plant, Golden, Colorado, November. [SRDB Ref ID: 159147]
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Rockwell International, 1985l, collection of collection of Routine Air Sample Results, Buildings 865 

and 886, Rocky Flats Plant, Golden, Colorado, December. [SRDB Ref ID: 159148]

Rockwell International, 1986a, collection of Building 886 radiological surveys, Rocky Flats Plant, 

Golden, Colorado, June. [SRDB Ref ID: 152308]

Rockwell International, 1986b, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, January. [SRDB Ref ID: 156882]

Rockwell International, 1986c, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, February. [SRDB Ref ID: 156882]

Rockwell International, 1986d, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, June. [SRDB Ref ID: 156887]

Rockwell International, 1986e, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, July. [SRDB Ref ID: 156889]

Rockwell International, 1986f, collection of Direct Alpha Count Air Sample Results, Buildings 865 and 

886, Rocky Flats Plant, Golden, Colorado, November. [SRDB Ref ID: 160325]

Rockwell International, 1986g, collection of Direct Alpha Count Air Sample Results, Buildings 865 and

886, Rocky Flats Plant, Golden, Colorado, December. [SRDB Ref ID: 156896]

Rockwell International, 1987a, collection of Routine Air Sample Results, Buildings 865 and 886, 

Rocky Flats Plant, Golden, Colorado, January. [SRDB Ref ID: 159149]

Rockwell International, 1987b, collection of Routine Air Sample Results, Buildings 865 and 886, 

Rocky Flats Plant, Golden, Colorado, February. [SRDB Ref ID: 159123]

Rockwell International, 1987c, collection of Routine Air Sample Results, Buildings 865 and 886, 

Rocky Flats Plant, Golden, Colorado, March. [SRDB Ref ID: 159150]

Rockwell International, 1987d, collection of Routine Air Sample Results, Buildings 865 and 886, 

Rocky Flats Plant, Golden, Colorado, May. [SRDB Ref ID: 159151]

Rockwell International, 1987e, collection of Routine Air Sample Results, Buildings 865 and 886, 

Rocky Flats Plant, Golden, Colorado, July. [SRDB Ref ID: 159152]

Rockwell International, 1987f, collection of Routine Air Sample Results, Buildings 865 and 886, Rocky 

Flats Plant, Golden, Colorado, August. [SRDB Ref ID: 159153]

Rockwell International, 1987g, collection of Routine Air Sample Results, Buildings 865 and 886, 

Rocky Flats Plant, Golden, Colorado, September. [SRDB Ref ID: 159156]

Rockwell International, 1987h, collection of Routine Air Sample Results, Buildings 865 and 886, 

Rocky Flats Plant, Golden, Colorado, December. [SRDB Ref ID: 159157]

Rockwell International, 1987i, Waste Stream Identification and Characterization, Overview, Rocky 

Flats Plant, Golden, Colorado, April 6. [SRDB Ref ID: 132654]

Rockwell International, 1987j, Waste Stream Identification and Characterization, Area 1, Rocky Flats 

Plant, Golden, Colorado, April 6. [SRDB Ref ID: 132655]
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Rockwell International, 1987k, Waste Stream Identification and Characterization, Radiochemical 

Compounds Data Report, Volume E-2, Rocky Flats Plant, Golden, Colorado, April 6. [SRDB 

Ref ID: 132656]

Rockwell International, 1987l, Waste Stream Identification and Characterization, Area 2, Volume 1, 

Rocky Flats Plant, Golden, Colorado, April 6. [SRDB Ref ID: 132658]

Rockwell International, 1987m, Waste Stream Identification and Characterization, Area 2, Volume 2, 

Rocky Flats Plant, Golden, Colorado, April 6. [SRDB Ref ID: 132657]

Rockwell International, 1987n, Waste Stream Identification and Characterization, Area 3, Rocky Flats 

Plant, Golden, Colorado, April 6. [SRDB Ref ID: 132659]

Rockwell International, 1987o, Waste Stream Identification and Characterization, Area 4, Volume 1, 

Rocky Flats Plant, Golden, Colorado, April 6. [SRDB Ref ID: 132660]

Rockwell International, 1987p, Waste Stream Identification and Characterization, Area 4, Volume 2, 

Rocky Flats Plant, Golden, Colorado, April 6. [SRDB Ref ID: 132661]

Rockwell International, 1987q, Waste Stream Identification and Characterization, Area 4, Volume 3, 

Rocky Flats Plant, Golden, Colorado, April 6. [SRDB Ref ID: 132662]

Rockwell International, 1987r, Waste Stream Identification and Characterization, Area 4, Volume 4,

Rocky Flats Plant, Golden, Colorado, April 6. [SRDB Ref ID: 132663]

Rockwell International, 1987s, Waste Stream Identification and Characterization, Area 4, Volume 5, 

Rocky Flats Plant, Golden, Colorado, April 6. [SRDB Ref ID: 132664]

Rockwell International, 1987t, Waste Stream Identification and Characterization, Area 4, Volume 6, 

Rocky Flats Plant, Golden, Colorado, April 6. [SRDB Ref ID: 132665]

Rockwell International, 1987u, Waste Stream Identification and Characterization, Inorganic 

Compounds Data Report, Volume A-1, Rocky Flats Plant, Golden, Colorado, April 6. [SRDB 

Ref ID: 132666]

Rockwell International, 1987v, Waste Stream Identification and Characterization, Volatile Compounds 

Data Report, Volume B-1, Rocky Flats Plant, Golden, Colorado, April 6. [SRDB Ref ID: 

132667]

Rockwell International, 1987w, Waste Stream Identification and Characterization, Volatile Compounds 

Data Report, Volume B-2 Rocky Flats Plant, Golden, Colorado, April 6. [SRDB Ref ID: 

132668]

Rockwell International, 1987x, Waste Stream Identification and Characterization, Volatile Compounds 

Data Report, Volume B-3, Rocky Flats Plant, Golden, Colorado, April 6. [SRDB Ref ID: 

132669]

Rockwell International, 1987y, Waste Stream Identification and Characterization, Volatile Compounds 

Data Report, Volume B-4, Rocky Flats Plant, Golden, Colorado, April 6. [SRDB Ref ID: 

132670]
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Rockwell International, 1987z, Waste Stream Identification and Characterization, Semi-Volatile 

Compounds Data Report, Volume C-1 Rocky Flats Plant, Golden, Colorado, April 6. [SRDB 

Ref ID: 132671]

Rockwell International, 1987aa, Waste Stream Identification and Characterization, Semi-Volatile 

Compounds Data Report, Volume C-2, Rocky Flats Plant, Golden, Colorado, April 6. [SRDB 

Ref ID: 132672]

Rockwell International, 1987ab, Waste Stream Identification and Characterization, Semi-Volatile 

Compounds Data Report, Volume C-3, Rocky Flats Plant, Golden, Colorado, April 6. [SRDB 

Ref ID: 132673]

Rockwell International, 1987ac, Waste Stream Identification and Characterization, Semi-Volatile 

Compounds Data Report, Volume C-4, Rocky Flats Plant, Golden, Colorado, April 6. [SRDB 

Ref ID: 132674]

Rockwell International, 1987ad, Waste Stream Identification and Characterization, Semi-Volatile 

Compounds Data Report, Volume C-5, Rocky Flats Plant, Golden, Colorado, April 6. [SRDB 

Ref ID: 132675]

Rockwell International, 1987ae, Waste Stream Identification and Characterization, Semi-Volatile 

Compounds Data Report, Volume C-6, Rocky Flats Plant, Golden, Colorado, April 6. [SRDB 

Ref ID: 132676]

Rockwell International, 1987af, Waste Stream Identification and Characterization, Pesticides/PCB’s 

Compounds Data Report, Volume D-1, Rocky Flats Plant, Golden, Colorado, April 6. [SRDB 

Ref ID: 132677]

Rockwell International, 1987ag, Waste Stream Identification and Characterization, Pesticides/PCB’s 

Compounds Data Report, Volume D-2, Rocky Flats Plant, Golden, Colorado, April 6. [SRDB 

Ref ID: 132678]

Rockwell International, 1987ah, Waste Stream Identification and Characterization, Radiochemical 

Compounds Data Report, Volume E-1, Rocky Flats Plant, Golden, Colorado, April 6. [SRDB 

Ref ID: 132679]

Rockwell International, 1987–1988, collection of Airhead & SAAM calibrations, Buildings 886 and 875, 

Rocky Flats Plant, Golden, Colorado. [SRDB Ref ID: 152336]

Rockwell International, 1988a, collection of Building 886 radiological surveys, Rocky Flats Plant, 

Golden, Colorado, January to October. [SRDB Ref ID: 152359]

Rockwell International, 1988b, collection of Airhead & SAAM calibrations, Buildings 886 and 875, 

Rocky Flats Plant, Golden, Colorado. [SRDB Ref ID: 152339]

Rockwell International, 1988c, collection of Air Sample Direct Alpha Counting System results, Building 

886, Rocky Flats Plant; January. [SRDB Ref ID: 156900]

Rockwell International, 1988d, collection of Air Sample Direct Alpha Counting System results, Building 

886, Rocky Flats Plant, Golden, Colorado, February 1988. [SRDB Ref ID: 156902]

Rockwell International, 1988e, collection of Air Sample Direct Alpha Counting System results, Building 

886, Rocky Flats Plant, Golden, Colorado, April. [SRDB Ref ID: 156907]
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Rockwell International, 1988f, collection of Routine Air Sample Results, Buildings 865 and 886, Rocky 

Flats Plant, Golden, Colorado, November. [SRDB Ref ID: 159159]

Rockwell International, 1988g, collection of Air Sample Results, Building 886, Rocky Flats Plant, 

Golden, Colorado, November. [SRDB Ref ID: 156161]

Rockwell International, 1989a, collection of Airhead & SAAM Calibrations, Buildings 886 and 875, 

Rocky Flats Plant, Golden, Colorado. [SRDB Ref ID: 152512]

Rockwell International, 1989b, collection of Airhead & SAAM Calibrations, Buildings 886 and 875, 

Rocky Flats Plant, Golden, Colorado. [SRDB Ref ID: 152513]

Rockwell International, 1989c, collection of Air Sample Results, Building 886, Rocky Flats Plant, 

Golden, Colorado, November. [SRDB Ref ID: 159163]

Rockwell International, 1989d, Organization and Top Issues, Rocky Flats Plant, Golden, Colorado, 

June 5. [SRDB Ref ID: 127205]

Rockwell International, 1990a, collection of Building 886 contamination surveys, Rocky Flats Plant, 

Golden, Colorado, January to June, October to December 1990. [SRDB Ref ID: 152515]

Rockwell International, undated, Rocky Flats Criticality Safety Experiment Run 2-8-170, Rocky Flats 

Plant, Golden, Colorado. [SRDB Ref ID: 157750]

Rothe, R. E., 2005, A Technically Useful History of the Critical Mass Laboratory at Rocky Flats, LAUR-05-3247, University of California, Los Alamos National Laboratory, Los Alamos, New 

Mexico, May. [SRDB Ref ID: 21358]

Sanchini, D., 1986–1988, excerpts from personal notebooks 2-4 and 6-7, Rockwell International, 

Rocky Flats Plant, Golden, Colorado, Golden, Colorado. [SRDB Ref ID: 143317]

Sanchini, D., 1987, “Mtg w/ Earl & Wayne, Earl’s Assessment of Wayne’s New Org,” Rockwell 

International, Rocky Flats Plant, Golden, Colorado, February 26. [SRDB Ref ID: 127209]

Sanchini, D., 1988a, “Discussion w/ M. Miles,” Rockwell International, Rocky Flats Plant, Golden, 

Colorado, February 24. [SRDB Ref ID: 127203]

Sanchini, D., 1988b, “Talk to R. Stazosteicki & Jim Knight,” Rockwell International, Rocky Flats Plant, 

Golden, Colorado, February 11. [SRDB Ref ID: 127204]

Sanchini, D., 1988c, “Mtg w/ John Auxier,” Rockwell International, Rocky Flats Plant, Golden, 

Colorado, March 15. [SRDB Ref ID: 127207]

Sanchini, D., 1988d, “Talk to John Auxier,” Rockwell International, Rocky Flats Plant, Golden, 

Colorado, March 7. [SRDB Ref ID: 127208]

Sanchini, D., 1988e, “Talk to John Auxier,” Rockwell International, Rocky Flats Plant, Golden, 

Colorado, June 6. [SRDB Ref ID: 127210]

Sanchini, D., 1988f, untitled letter to A. E. Whiteman (U.S. Department of Energy, Rocky Flats Area 

Office), Rockwell International, Rocky Flats Plant, Golden, Colorado, December 19. [SRDB 

Ref ID: 127212]
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Sanchini, D., 1988–1989, excerpts from personal notebook 8, Rockwell International, Rocky Flats 

Plant, Golden, Colorado. [SRDB Ref ID: 143318]

SC&A (S. Cohen and Associates), 2007a, A Focused Review of Operations and Thorium Exposures 

at The Dow Chemical Company Madison Plant, SCA-SEC-TASK5-0057, Vienna, Virginia, 

August. [SRDB Ref ID: 53614]

SC&A (S. Cohen and Associates), 2007b, “Internal Dose Issues”, Section 6.0 of RFETS SEC Petition 

Review, Working Draft, Vienna, Virginia, February 15. [SRDB Ref ID: 39990]

Schuske, C. L., 1977, Total Fissions Produced by Critical Approach Experiments Done at the Rocky 

Flats Plant Nuclear Facility from September, 1965 to June 1, 1977, letter to W. R. Stratton 

(University of California), Atomics International, Rocky Flats Plant, Golden, Colorado, June 3. 

[SRDB Ref ID: 157872]

Scientech, 1989, Independent Criticality Safety Assessment, June to September, Rocky Flats Plant, 

Golden, Colorado, October 4. [SRDB Ref ID: 133226]

Sebelius, K., 2013, HHS Designation of Additional Members of the Special Exposure Cohort Under 

the Energy Employees Occupational Illness Compensation Program Act of 2000, Designating 

a Class of Employees from Rocky Flats Plant, Golden, Colorado, U.S. Department of Health 

and Human Services, Office of the Secretary, Washington, D.C., December 12. [SRDB Ref 

ID: 166855]

Silverstein, L., 1957, “Mg-Th Fires,” memorandum to B. Saunders, The Dow Chemical Company, 

Madison, Illinois, July 31. [SRDB Ref ID: 39989]

Silverstein, L., 1958, “Thorium Dust Samples Taken During HK31 Sanding,” memorandum to C.A. 

Mitchell, The Dow Chemical Company, Madison, Illinois, April 16. [SRDB Ref ID: 39905]

Simonson, D. P., 1989, “U. S. Attorneys Request for Documents,” memorandum to AMs, Division 

Directors and Branch Chiefs, U.S. Department of Energy, Rocky Flats Office, Golden, 

Colorado. [SRDB Ref ID: 127262]

SRGA (Stephen Ray Green and Associates), 2008, “Meeting Fifty-Six, Advisory Board on Radiation 

and Worker Health, Vol. II, Day Two, ABRWH Board Meeting,” transcript, St. Louis, Missouri, 

June 25. [SRDB Ref ID: 53615]

Swenson, B. A., 1995, Are You Prepared To Survive An FBI Raid At Your Facility?, Kaiser-Hill 

Company, Rocky Flats Plant, Golden, Colorado, August 28. [SRDB Ref ID: 122696]

Ulsh, B. A., B. L. Rich, M. H. Chew, R. L. Morris, M. Sharfi, and M. R. Rolfes, 2008, “Establishing 

Bounding Internal Dose Estimates for Thorium Activities at Rocky Flats,” Health Physics, 

volume 95, number 1, pp. 81–88. [SRDB Ref ID: 122613]

Various, 1952–1982, collection of bioassay records for former Rocky Flats Plant worker. [SRDB Ref 

ID: 25473] 

Various, 1955–1976, collection of bioassay records for former Rocky Flats Plant workers. [SRDB Ref 

ID: 31256]

Various, 1958–2003, collection of bioassay records for former Rocky Flats Plant worker. [SRDB Ref 

ID: 24765] 
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Various, 1954–1996, collection of bioassay records for former Rocky Flats Plant worker. [SRDB Ref 

ID: 26707]

Various, 1961–1975, collection of information on MgTh alloy. [SRDB Ref ID: 108264]

Various, 1965–1987, collection of records for former Rocky Flats Plant worker. [SRDB Ref ID: 

71935]

Various, 1982–2005, collection of records for former Rocky Flats Plant worker. [SRDB Ref ID: 

25528]

Vejvoda, E., 2005, Summary of Rocky Flats Plant Waste Buried in the Subsurface Disposal Area, 

ICP/EXT-04-00717, Rev. 0, Los Alamos Technical Associates, Aurora, Colorado, April, pp. 80-

84. [SRDB Ref ID: 33009]

Walraven, D. J., 1991, “Lung Counting Software - DJW-001-92,” letter to R. Hume (Livermore, 

California), EG&G Rocky Flats, Rocky Flats Plant, Golden, Colorado, January 10. [SRDB Ref 

ID: 133702]
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GLOSSARY

acute exposure

Radiation exposure to the body delivered in a short period. See chronic exposure. 

alpha particles

See alpha radiation. 

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

chronic exposure

Radiation dose to the body delivered in small amounts over a long period (e.g., days or years). 

See acute exposure. 

curie

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

detection limit (lower)

See limit of detection. 

dose

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of roentgens, rad, rep, or grays.

dosimetry

Measurement and calculation of internal and external radiation doses.

element

One of the known chemical substances in which the atoms have the same number of protons. 

Elements cannot be broken down further without changing their chemical properties. 

Chemical symbols for the elements consist of either a single letter or a combination of letters, 

some of which descend from the Latin names [e.g., Au from aurum (gold), Fe from ferrum

(iron)]. This glossary indicates elements by their names. Specific isotopes appear as their 

standard chemical symbols with the number of protons and neutrons in the nucleus. For 

example, the isotope of uranium that contains 92 protons and 143 neutrons can appear as 

235U, U-235, or uranium-235. See periodic table of the elements and radioactive isotope. 

exposure

(1) In general, the act of being exposed to ionizing radiation. See acute exposure and chronic 

exposure. (2) Measure of the ionization produced by X- and gamma-ray photons in air in units 

of roentgens.
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extremities

Portion of the arm from and including the elbow through the fingertips and the portion of the 

leg from and including the knee and patella through the toes.

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus.

gamma ray

See gamma radiation.

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. See 

alpha radiation, beta radiation, gamma radiation, neutron radiation, photon radiation, and X-ray 

radiation. 

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties. See element. 

limit of detection

Minimum level at which a particular device can detect and quantify exposure or radiation. Also

called lower limit of detection and detection limit or level.

maximum permissible lung burden (MPLB)

Historical occupational limit on the amount of a radionuclide present in the systemic body at 

the end of 50 years as a result of being exposed at the maximum permissible concentration for 

50 working years.

minimum detectable amount (MDA)

Smallest amount (activity or mass) of an analyte in a sample that can be detected with a 

probability β of nondetection (Type II error) while accepting a probability α of erroneously 

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

(Type I error).

neutron (n)

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen. See element. 

neutron radiation

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons. See neutron. 
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nuclide

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

periodic table of the elements

Arrangement of the chemical elements in order of increasing atomic number from left to right 

and by similar chemical properties vertically. Elements of similar properties occur one under 

the other, which yields groups or families of elements. 

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz.

photon radiation

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays.

radioactive isotope

Natural or synthetic form of an atom that emits radioactivity when it decays. See isotope. 

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. See ionizing radiation. 

radioactive

Of, caused by, or exhibiting radioactivity.

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei.

radionuclide

Radioactive nuclide. See radioactive and nuclide. 

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

site returns

At Rocky Flats, weapons components returned from other sites for disassembly and recovery 

of materials.

whole-body dose

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 

10 millimeters (1,000 milligrams per square centimeter). Also called penetrating dose. 

See dose. 
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X-ray radiation

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.
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ATTACHMENT A 

MINIMUM DETECTABLE AMOUNTS FOR URINALYSIS

TABLE OF CONTENTS

SECTION TITLE PAGE

A.1 Introduction ............................................................................................................................ 70

A.2 MDA Methodology.................................................................................................................. 70

A.3 History of Methods ................................................................................................................. 71

A.4 Assessment of MDA............................................................................................................... 77

A.5 MDA Values........................................................................................................................... 81

A.6 Discussion ............................................................................................................................. 87

LIST OF TABLES

TABLE TITLE PAGE

A-1 Method codes ........................................................................................................................72

A-2 Correlation of method code and analyte................................................................................. 72

A-3 Values of tolerance and reporting levels................................................................................. 72

A-4 Sample volumes for routine 24-hour urine samples................................................................ 77

A-5 Detector background for gas flow proportional counters......................................................... 78

A-6 Median and 95th-percentile blank count rates........................................................................ 78

A-7 Efficiencies of alpha-counting detectors ................................................................................. 79

A-8 Recovery fractions used in MDA assessments....................................................................... 79

A-9 Fractions of alphas absorbed in residue at the 95th percentile............................................... 80

A-10 Gross alpha calibration factor................................................................................................. 81

A-11 Values of variables and MDA for plutonium for median conditions ......................................... 82

A-12 Values of variables and MDA for plutonium for extreme conditions ........................................ 82

A-13 MDA for plutonium for one, two, or three extreme conditions ................................................. 82

A-14 Values of variables and MDA for EU for median conditions.................................................... 83

A-15 Values of variables and MDA for EU for extreme conditions .................................................. 83

A-16 MDA for EU for one, two, or three extreme conditions............................................................ 84

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

extreme conditions.................................................................................................................84

A-18 Values of variables and MDA for americium for median conditions ........................................ 84

A-19 Values of variables and MDA for americium for extreme conditions ....................................... 85

A-20 Values of the MDA for americium for one, two, or three extreme conditions........................... 85

A-21 Values of variables and MDA for gross alpha measurements for median conditions .............. 86

A-22 Values of variables and MDA for gross alpha measurements for extreme conditions............. 86

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

conditions...............................................................................................................................86
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ATTACHMENT A

MINIMUM DETECTABLE AMOUNTS FOR URINALYSIS (continued)

A.1 INTRODUCTION

Urinalysis was 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 mass 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 after an actual or suspected intake. 

Routine urine samples were typically 24-hour excretions, either one continuous 24-hour period (but 

not taken at the RFP site) or two 12-hour periods. Special urine samples could be 24-hour 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. The analytes of interest are plutonium, americium, EU, and DU. In addition, 

RFP analyzed for gross alpha using a nonspecific analysis 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 found and analyzed to obtain the 

information necessary 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):

(A-1) ( ) ( 1 Δ 2Δ 2 3 K BB ks0 ) MDA

KT

+ ++ =

where

B = the total count of the appropriate blank

ΔB = the maximum expected fractional systematic error bound in the appropriate blank

K = calibration factor

ΔK = the maximum fractional systematic error bound in the calibration factor K

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

s0 = the standard deviation in the net count of a sample with no additional analyte:

2

2 B0

0 B1 2

s

s s

m

= +
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ATTACHMENT A

MINIMUM DETECTABLE AMOUNTS FOR URINALYSIS (continued)

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

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 that were 

processed in those plants as well for the urinalysis procedures that were 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. 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, 

Rocky Flats Site, Building 991” (DOE 2011a). 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 for the 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 that was 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. 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.

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 (see Figures C-1 to C-3 in 

Attachment C). 

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 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 
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Table A-1. Method codes.

Code Meaning

A Fluorimeter, reported in μg/L 1952–1956 and 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

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

Analyte Method code

DU A, B1 (starting 05/01/64)

EU B1

Plutonium C1, C2

Gross alpha B2, B3, D

exceptions. Table A-3 lists the values of the tolerance levels. The tolerance levels were noted as the 

“working MDL” (minimum detection level) 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.

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

Analyte Tolerance level Reporting level

DU 58 μg/24 hr ≥5.8 μg/24 hr

EU 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. 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.

For plutonium, the reporting and recording level was ≥0.2 dpm/24 hr for 1962 to April 6, 1970 based 

on an examination of the CEDR database. 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 depending on how far back it was applied.

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 (ORAUT 2003). 
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The general method for data analysis for alpha-counting procedures (1952 to 1971) was:

( ) (A-2) ( Det Blk )( ) dpm/24-hr sample

ε

C V T A B B

Activity R

− − =

where

C = total count

T = count time (min)

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/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 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 that was used in the data reduction was generally tabulated in the urinalysis data logs.

The ratio V/A is a volume adjustment factor that was used for two purposes. If the entire sample was 

not analyzed, this ratio normalized the result from the volume of the analyzed aliquant 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. 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. 

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 that were 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 deposited activity (dpm).

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 εR was frequently combined, especially in the 1950s. In the 1960s, the term 1/εR 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.
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The general method for the mass measurements of uranium using the fluorimeter (1953 to 1964) was:

( ) (A-3) Blk μg/24-hr sample S B Mass

K

− =

where

S = signal reading of the sample aliquant

BBlk = signal reading of the blank

K = constant/V; 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 [position redacted] 

from [date redacted] to [date redacted], [name redacted], in 1992 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 minutes. A spike sample and a reagent 

blank sample were processed with the worker samples, sometimes with each batch and 

sometimes less frequently. The result of the spike sample might 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) might 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 (ORAUT 2003). 

• 1961 to 1962. Starting on December 13, 1961, a thenoyltrifluoroacetone (TTA) extraction step 

was added to the carrier precipitation method to improve the specificity of the process to 

isolate plutonium (ORAUT 2003). 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 specific to 

plutonium. 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 onethird 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 (ORAUT 2003). 

• 1978 to 1993. By 1978, all counting systems had been converted to the PHA system, and all 

plutonium samples were processed with internal tracers. The recovered fraction of the internal 
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tracer for that sample was applied in the analysis of the result for that sample. The acceptable 

range of the fractional tracer recovery was 0.10 to 1.10. The result of a sample was 

invalidated if the recovery was outside the acceptable range (ORAUT 2003). In 1990, the 

acceptable recovery range was changed to 0.35 to 1.10 (ORAUT 2003). The count time of 

720 minutes was used for all samples. A batch blank continued to be processed and generally 

was used in the data analysis unless suspected to have been contaminated excessively (a 

subjective decision). In 1985, the blank method was modified. The value of the blank that 

was 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 analyzed sample (aliquant) was 800 mL if the volume 

of the sample was greater than 800 mL or, if the volume of the sample was less than 800 mL, 

the entire sample. 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 After. Upgrades to procedures occurred in 1993 to achieve a process MDA less 

than or equal to 0.020 dpm/sample (ORAUT 2003). Count time was increased to 1,400 

minutes. The entire sample was analyzed so that the volume fraction was unity for all 

samples. In addition, a contract was established with a commercial bioassay laboratory, with 

a requirement that an MDA of ≤0.02 dpm/sample be achieved. In 1997, the onsite bioassay 

laboratory was shut down.

Americium (1963 and After) 

Except for the details of the chemistry, the process for americium was similar to that for plutonium. A 

solvent extraction process, specific for americium, was first used in 1963 (ORAUT 2003). 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-hour 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 in this period were 30, 40, 60, 70, 90, 120, and 

150 minutes.

From 1952 to 1955, one aliquant per sample was used. In 1960, a second aliquant was processed if 

the result of the first aliquant was ≥7 dpm/24-hr sample. If the second result was within a specified 

range of the first result, the average of the two results was recorded and reported. If the second result 

was out of the specified range, a third aliquant was processed, and the average of the two results that 

best confirmed each other was used. If that average was less than the reporting level of 

8.8 dpm/24 hr, the result was recorded and reported as background. From 1961 to 1971, two 

aliquants were routinely processed for each urine sample, with a third aliquant (1961 to 1969) 

processed if the spread of the results of the first set was outside the specified range. The recording 
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and reporting logic was the same as that for 1960. From 1964 to 1971, the recording and reporting 

limit appears to have been ≥20 to 28 dpm/24-hr sample, depending on the volume of the sample 

(ORAUT 2003). 

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) (ORAUT 2003). 

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-hour 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 that 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 of 

5.8 μg/24-hr sample. Otherwise, the result was reported as background.

After April 1964, the urine sample was analyzed using the electroplating procedure described above 

for EU, and the results were reported in dpm/24-hr sample (or background) (ORAUT 2003). 

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 tributyl 

phosphate (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 (ORAUT 2003). 

In all methods, the entire urine sample was processed, and the final extract was evaporated on a 

planchet and counted on the gas-flow proportional counter. Counting time was typically 150 minutes, 

although from 1952 to 1955 count times of 55, 60, and 75 minutes, and in 1971 40 and 60 minutes, 

were also used.

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

system to determine whether to credit the result to EU or to plutonium, or a portion to each. The 

default assumption 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 assumption was to credit 

the result to plutonium.
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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 

consideration of 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 nonnegative 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 Δ value of ΔK. 

Table A-4 lists sample volumes for routine 24-hour urine samples.

Table A-4. Sample volumes (mL) for 

routine 24-hour urine samples.

Percentile Volume

5th 700

Median 1,350

95th 1,750

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.

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 
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of 150 minutes. To have a coherent dataset, only background count data for samples counted for 150 

minutes were extracted from the urine data logs.

Table A-5. Detector background (cpm) for gas flow proportional counters.

Date 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 that was used to measure the detector 

background, but the count time was probably 150 minutes or longer. For the purpose of assessing 

the MDA, the composite average is used for the value of the detector background count rate BDet = 

0.056 cpm with the standard deviation sDet = 0.017 cpm for all alpha-counting methods (except for 

americium) and for all sample count times. 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. For americium, the values for the 1960s are used because the americium process was not 

implemented in the 1950s. This exception was made for americium because the detector 

backgrounds for the 1950s did not apply.

The blank count rate was method-specific, and the application of the blank in the data analysis was 

variable between methods and within a method over time. A complication that was intermittent, but 

persistent, was laboratory contamination artifacts that were introduced into blanks and worker 

samples. These artifacts caused false positives from a worker exposure viewpoint but real positives 

from a detection viewpoint.

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 (cpm). 

Analyte Period Median 95th percentile

Plutonium 1952–1971 0.06 0.28

EU 1952–1971 0.05 0.22

DU 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 in 2003 (the initial author of this TBD; see Section 5.7). 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. 

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:

(A-4) Blk

Blk

B T

s

T =
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The values for B, sB0, sB1, and s0 in the MDA equation (A-1) are derived from the detector background 

and blank values:

B TB B = + ( ) Det Blk (A-5)

(A-6) 2 2

B0 Det Blk s Ts s = +

(A-7) B1 s B =

(A-8) 2 2

0 B1 B0 s ss = +

For some analytes (EU, DU) and periods, the detection decision was based on the average of two 

aliquants. In this case, the value of s0 for the average of two aliquants is equal to the value of s0 for 

one aliquant divided by the square root of 2. 

The value of ΔB 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. This approach 

is consistent with this section under the General Considerations subsection. 

Table A-8 lists the recoveries that were 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.

Table A-8. Recovery fractions used in MDA assessments.

Analyte Period Median 5th percentile

Plutonium 1952–1962 0.57 0.25

Plutonium 1963–1971 0.67 0.28

EU 1952–1971 0.60 0.21

DU 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
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The recovery values are based on incomplete datasets 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 that was 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 November 1, 1965, the values are based on a complete set for that period.

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.

The volume adjustment factor V/A is incorporated into the calibration factor K as the reciprocal 

1/(V/A), s0 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 5thpercentile volume), and Vf = 0.58 (ORAUT 2003). 

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 (ORAUT 2003). 

The absorption of the alpha particles in the residue that was 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 that are 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. 

Table A-9. Fractions of alphas absorbed in residue at the 

95th percentile. 

Analyte Period 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

EU (electrodeposited) 1952–1971 0.05 0.95

DU (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 discernible 

from the databases and reports of urinalysis results that are readily accessible. 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. For the purpose of the 

MDA assessment, dose reconstructors should use the value of Fa for the evaporation process.
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MINIMUM DETECTABLE AMOUNTS FOR URINALYSIS (continued)

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 at the 

time.

The calibration factor K is the product of ε, R, Vf, and Fa: 

ε a (A-9) K RV Ff =

The values of ΔB and ΔK are considered to be zero. No evidence of a systematic bias in the 

background or the calibration factor was discerned by the author. Therefore, ΔB and ΔK were set 

equal to zero.

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 s0 is equal to the value of sB0 

multiplied by the square root of 2:

s s 0 B0 = 2 (A-10) 

The value of sB0 was determined from a review by R. 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 μg U/24-hr sample (see 

Equation A-3). 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 (ORAUT 2003). The factor 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.

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 ΔB and ΔK 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.
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ATTACHMENT A

MINIMUM DETECTABLE AMOUNTS FOR URINALYSIS (continued)

Plutonium

The MDA for plutonium is assessed for the median condition and for the extreme (5th- or 95thpercentile) condition for the blank, the recovery, the volume factor Vf, and the alpha transmission 

factor Fa, individually and in combination. A count time of 150 minutes is used for all assessments.

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.

MDA

Period s0 ε R Vf Fa

(dpm/24-hr 

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 s0 ε R Vf Fa

MDA

(dpm/24-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 s0 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-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 s0 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 s0, R s0, Vf s0, 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
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ATTACHMENT A

MINIMUM DETECTABLE AMOUNTS FOR URINALYSIS (continued)

MDA (dpm/24-hr sample) for three extreme conditions

Period s0, R, Vf s0, R, Fa s0, 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

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 minutes is used for MDA assessments from 1952 

to 1963.

For 1964 to 1969, the count time of 30 minutes is used and, for 1970 to 1971, the count time of 

40 minutes is used. For 1952 to 1959, the value of s0 is calculated for one aliquant, and for 1960 to 

1971 the value of s0 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 s0 ε R Vf Fa

MDA

(dpm\24-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

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 s0 ε R Vf Fa

MDA

(dpm\24-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 s0 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-17 lists the values of variables and MDAs for fluorimetric measurements of DU for median 

and extreme conditions.
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ATTACHMENT A

MINIMUM DETECTABLE AMOUNTS FOR URINALYSIS (continued)

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

MDA (dpm/24-hr sample) for one extreme condition

Period s0 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 s0, R s0, Vf s0, 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 s0, R, Vf s0, R, Fa s0, 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

Table A-17. Values of variables and MDA for fluorimetric measurements 

of DU for median and extreme conditions.

Period sB0

Median 

K

Extreme 

K

Median

(μg/24-hr 

sample)

Extreme

(μg/24-hr 

sample)

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

Americium

The MDA for americium is assessed for the median condition and for the extreme (5th- or 95thpercentile) condition for the blank, the recovery, the volume factor Vf, and the alpha transmission 

factor Fa, individually and in combination. A count time of 150 minutes is used for assessments from 

1963 to 1970. In 1971, the typical (and minimum) count time is 60 minutes.

Table A-18 lists the MDA to two significant figures.

Table A-18. Values of variables and MDA for americium for 

median conditions.

Period s0 ε R Vf Fa

MDA

(dpm/24-hr 

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
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ATTACHMENT A

MINIMUM DETECTABLE AMOUNTS FOR URINALYSIS (continued)

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.

Table A-19. Values of variables and MDA for americium for 

extreme conditions.

Period s0 ε R Vf Fa

MDA

(dpm/24-hr 

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 s0 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 s0 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 s0, R s0, Vf s0, 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 s0, R, Vf s0, R, Fa s0, Vf, Fa R, Vf, Fa

1963 3.0 2.5 1.7 2.8

1964–1965 3.8 3.2 2.1 3.5

1965–1970 3.8 3.2 1.8 3.5

1971 6.2 5.2 2.9 5.7

Gross Alpha

The MDA for gross alpha measurements is assessed for the median condition and for the extreme 

(5th- or 95th-percentile) condition for the blank, the recovery, the volume factor Vf, and the alpha 

transmission factor Fa, individually and in combination. A count time of 55 minutes is used for 1952, 

75 minutes for 1953 to 1959, and 150 minutes for 1960 to 1971 for assessments of the MDA for both 

the median and extreme conditions, except for 1971, when a count time of 40 minutes is also used for 

the extreme condition. See Table A-21.
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ATTACHMENT A

MINIMUM DETECTABLE AMOUNTS FOR URINALYSIS (continued)

Table A-21. Values of variables and MDA for gross alpha 

measurements for median conditions.

Period s0 ε R Vfee Fa

MDA

(dpm/24-hr 

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 s0 ε R Vf Fa

MDA

(dpm/24-hr 

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 s0 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 (dpm/24-hr sample) for gross alpha measurements for one, two, or 

three extreme conditions.

One extreme condition

Period s0 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
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ATTACHMENT A

MINIMUM DETECTABLE AMOUNTS FOR URINALYSIS (continued)

Two extreme conditions

Period s0, R s0, Vf s0, Fa R, Vf R, Fa Vf, Fa

1952 3.9 2.8 1.8 4.3 2.8 2.0

1953 3.3 2.4 1.5 3.6 2.3 1.7

1954–1959 2.9 2.1 1.4 3.2 2.1 1.5

1960–1962 2.0 1.5 0.96 2.3 1.5 1.1

1963 2.0 1.5 1.2 2.3 1.9 1.4

1964–1971 2.6 1.9 1.5 2.8 2.3 1.7

1971 (T = 40 min) 5.2 3.8 3.1 5.8 4.8 3.5

Three extreme conditions

Period s0, R, Vf s0, R, Fa s0, 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.
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS

TABLE OF CONTENTS

SECTION TITLE PAGE

B.1 Introduction ............................................................................................................................ 89

B.2 MDA Methodology.................................................................................................................. 89

B.3 History of Counting Systems and Procedures ........................................................................ 90

B.4 Assessment of MDA............................................................................................................... 93

B.5 Values of the Variables, NaI(Tl) and Phoswich Detector Systems, 1969................................ 95

B.6 Values of the Variables, Ortec Germanium Detector Systems, 1976...................................... 96

B.7 Values of the Variables, PGT I Germanium Detector Systems, 1978 ..................................... 97

B.8 Values of the Variables, PGT II Germanium Detector Systems, 1979 .................................... 99

B.9 Values of the Variables, PGT Organ Pipe Germanium Detector Systems, 1985 .................. 100

B.10 Values of the Variables, EG&G Organ Pipe Germanium Detector Systems, 1985 ............... 100

B.11 MDA for RFP Plutonium....................................................................................................... 102

LIST OF TABLES

TABLE TITLE PAGE

B-1 Discontinuity factors...............................................................................................................93

B-2 Calibration factors for the eight-detector Ortec germanium system ........................................ 97

B-3 Calibration factors for the eight-detector PGT I germanium system........................................ 98

B-4 Calibration factors for the eight-detector PGT II germanium system....................................... 99

B-5 Values of variables for the PGT organ pipe germanium detector system ............................. 100

B-6 Calibration factors for the six-detector PGT organ pipe germanium detector system............ 100

B-7 Values of variables for the EG&G organ pipe germanium detector system .......................... 101

B-8 Calibration factors for the six-detector EG&G organ pipe germanium detector system......... 101

B-9 MDA conversion factors for values of ppm 241Am ................................................................. 102

B-10 Americium-241 ingrowth in plutonium................................................................................... 102

B-11 Americium-241 MDA for in vivo lung counts......................................................................... 104
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

B.1 INTRODUCTION

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 that were 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 L X-rays. Instead, the 

59.5-keV gamma photon from 241Am was used as a surrogate. Americium-241 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):

(B-1) ( ) ( 1 Δ 2Δ 2 3 K BB ks0 ) MDA

KT

+ ++ =

where

ΔK = the maximum fractional systematic error bound in the calibration factor K

ΔB = the maximum expected fractional systematic error bound in the appropriate blank

B = the total count of 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

K = calibration factor

s0 = the standard deviation in the net sample count of a subject with no additional analyte

(B-2) 2

2 B0

o B1 2

s

s s

m

= +

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

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 

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.
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

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 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 (Mann 1968, p. 60; 1969, p. 10): 

• Room A, built in 1964, operational in 1965;

• Room B, built in 1968, operational in 1969; and

• Room C, built in 1975, operational in 1976. 

Each room was equipped with a detector system. When a new detector system was implemented, 

the previous system was usually maintained as a 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 situation, the detector system 

is identified in the record for each lung count.

1964 to 1968 (Author unknown 2002; Boss and Mann 1967) 

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 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 Radiation-Equivalent Man, Absorption phantom from Alderson Research 

Laboratories. No adjustment was made for CWT.

1969 to 1976 (Author unknown 2002)

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.

Changes during this period include: 

• The ROI of the 59.5-keV photopeak of 241Am was expanded.

• 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.

• The standard count time was changed to 2,000 s (1,000 s for expedited counts).
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

• 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):

• Subjects were characterized by an index I equal to the ratio of the subject’s weight W in 

pounds divided by twice the subject’s height H in inches.

• A population of at least 20 known cold (unexposed) subjects of a diversity of indices was 

counted to generate a dataset of net count rate versus index.

• A curve fit to the dataset generated a prediction equation with the index as the variable.

• The subject’s index was used to determine the predicted net count rate for the subject.

This approach was applied separately for the right chest, the left chest, and the liver/gut.

In 1973, a phoswich detector system [a detector with a primary scintillation NaI(TI) layer backed by a 

CsI 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 using 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 RFP used was the five-point moving average of 

the last five counts.

Starting in 1969 (RFP 1965–1986), 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.

1976 to 1985 (Author unknown 2002)

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 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. To maintain a minimum functional system, a hybrid system 

consisting of two arrays of different characteristics was frequently used.
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

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, and

• 1980. Second array, PGT II detectors, 18 cm2 per detector.

Other changes in this period were:

• 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 = 2I – 0.1).

• Calibration was accomplished using a Masonite phantom from 1976 to 1978.

• Calibration was accomplished using the LLNL phantom starting in 1979.

• 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 an ROI 

in the range of 65 to 72 keV.

1985 to 1995 (Author unknown 2002; Walraven 1991) 

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 on the right and two on the left. The full system was seven 

detectors, four on the right and three on the left, although the routine system generally consisted of 

six, 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 on the right and two on the left or three on each side.

The germanium system implementation timeline was:

• 1985. PGT organ pipe detectors, 20 cm2 per detector, and

• 1991. EG&G Ortec organ pipe detectors, 20 cm2 per detector. 

No other significant changes were made during this period.

1995 to 2005 (KHC 2000b)

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 previously, 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.
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

Another significant change (KHC 2000b) was the equation to determine CWT. ABACOS-Plus 

incorporates the equation developed at LLNL:

(B-3) 1 973 (cm) 2 0038 . W CWT . H = −

where

W = subject’s weight (pounds)

H = subject’s height (inches)

The effect of this change is an adjustment factor given by:

(B-4) 0 635 0 5364e . I CWT Adjustment Factor . =

This adjustment factor is a multiplier to the activity of 241Am, detected using the 59.5-keV gamma, for 

all previous detector systems at RFP. Equation B-4 can also be applied as a divisor to 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 that affected the interpretation of the raw 

data (and therefore the MDA), were identified through 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 for each discontinuity were 

then applied as divisors to the calibration factor 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 + ΔK) in 

Equation B-1. Table B-1 lists the discontinuity factors.

Table B-1. Discontinuity factors.

Year Discontinuity Factor

1995 New CWT method, Index = 0.90 0.95

1995 Index = 1.35 1.26

1995 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

Values of the Variables, 1964 to 1968

The minimum system was one NaI(Tl) detector over the left chest.

Count time T = 20 MLT or 40 MLT (Various 1952–1982, p. 360; Various 1954–1996, p. 121; 

Various 1958–2003, p. 115) 
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

The appropriate blank B was the net subject 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

ΔB = 0.2, estimated as the upper bound for this method

The value of s0 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.

sB12 = Total subject count + R = B + 2R

= 1,600 for T = 20 MLT

= 3,200 for T = 40 MLT

sB02 = B

= 600 for T = 20 MLT

= 1,200 for T = 40 MLT

m = 1 

s0 = 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 

and incorporating the discontinuity factors (Equation B-4, 1.30, and 1.45) is given by:

(B-5) 0 2359 2 0 1 ( )

0 635

55 13e

e

. I.

. I

. K

− −

=

For Equation B-5 and for subsequent equations of the calibration factor K, the calibration factors were 

normalized to Lawrence Livermore National Laboratory (LLNL) phantom, also called the Lawrence 

Livermore Torso Phantom and described in KHC (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.”

The calibration factor for the system with only one detector over the left portion of the chest is given 

by Equation B-5 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
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

Because K is normalized to the calibration with the LLNL phantom and the discontinuity factors are 

incorporated into K, the value of ΔK is taken to be zero. Because the term (1 + ΔK) in Equation B-1 is 

a multiplier to the MDA and because the value of ΔK 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

ΔB = 0.2, estimated as the upper bound for this method

R = 1,000 for T = 20 MLT

R = 2,000 for T = 40 MLT

sB12 = Total subject count + R = B + 2R

 = 3,200 for T = 20 MLT

 = 6,400 for T = 40 MLT

sB02 = B

 = 1,200 for T = 20 MLT

 = 2,400 for T = 40 MLT

m = 1 

s0 = 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, NaI(Tl) AND PHOSWICH DETECTOR SYSTEMS, 1969

The standard system was two detectors 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 (RFP 1965–2005, pp. 22, 30, 83, 368, 803, 811, 817) 

The appropriate blank was the net subject 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 

ΔB = 0 for the NaI(Tl) detector system

ΔB = 0.1, estimated for the phoswich detector system, because the system was less stable 

than the NaI(Tl) detector system

The value of s0 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 

sB12 = Total subject count + R = B + 2R
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

= 2,767 for T = 1,000 s 

= 5,533 for T = 2,000 s 

sB02 = (0.1B)

2

= 12,100 for T = 1,000 s 

= 48,400 for T = 2,000 s 

m = 1 

s0 = 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-4 and 1.30) is given by:

(B-6) 0 2359 2 0 1 ( )

0 635

79 94e

e

. I.

. I

. K

− −

=

Which yields the following results:

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, ORTEC GERMANIUM DETECTOR SYSTEMS, 1976

The standard system was two arrays, each array with four detectors 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.

ΔB = 0 

m = 8 

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.
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

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-4 and 1.30 (for pre1979 systems)] is given by:

(B-7) 0 3398 2 0 1 ( )

0 635

24 12e

e

. I.

. I

. K

− −

=

and, for Ortec systems 1979 and after:

(B-8) 0 3398 2 0 1 ( )

0 635

31 36e

e

. I.

. I

. K

− −

=

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 (K) for the eightdetector Ortec germanium system.

Index Pre-1979 1979

0.90 7.64 9.94

1.35 4.23 5.50

1.80 2.34 3.04

B.7 VALUES OF THE VARIABLES, PGT I GERMANIUM DETECTOR SYSTEMS, 1978

The PGT I germanium system is basically the same as the Ortec germanium system.

ΔB = 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)
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

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-4 and 1.30 (for pre1979 systems)] is given by:

(B-9) 0 3292 2 0 1 ( )

0 635

34 09e

e

. I.

. I

. K

− −

=

and, for PGT I systems 1979 and after:

(B-10) 0 3292 2 0 1 ( )

0 635

44 318e

e

. I.

. I

. K

− −

=

For the minimum system of five detectors, adjust the calibration factor by multiplying by 0.625 (5/8).

Table B-3 lists calibration factors for the PGT I germanium detector system.

Table B-3. Calibration factors (K) for the eightdetector PGT I germanium system. 

Index Pre-1979 1979 →

0.90 11.00 14.30

1.35 6.15 7.99

1.80 3.43 4.46
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

B.8 VALUES OF THE VARIABLES, PGT II GERMANIUM DETECTOR SYSTEMS, 1979

The PGT II germanium system is basically the same as the Ortec and PGT I systems.

ΔB = 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

The 241Am calibration factor K for two arrays with a total of eight detectors (incorporating 

Equation B-4), is given by:

(B-11) 0 3579 2 0 1 ( )

0 635

38 65e

e

. I.

. I

. K

− −

=

For the minimum system of five detectors, adjust the calibration factor by multiplying by 0.625 (5/8).

Table B-4 lists calibration factors for the PGT II germanium detector system.

Table B-4. Calibration factors (K) for the 

eight-detector PGT II germanium system.

Index Calibration factor (K)

0.90 11.88

1.35 6.47

1.80 3.52
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

B.9 VALUES OF THE VARIABLES, PGT ORGAN PIPE GERMANIUM DETECTOR SYSTEMS, 

1985

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.

ΔB = 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. 

Variable 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-4) 

is given by:

(B-12) 0 2946 2 0 1 ( )

0 635

34 32e

e

. I.

. I

. K

− −

=

Table B-6 lists calibration factors for the PGT organ pipe germanium detector system.

Table B-6. Calibration factors (K) for the 

six-detector 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, EG&G ORGAN PIPE GERMANIUM DETECTOR SYSTEMS, 

1985

The EG&G organ pipe germanium system is basically the same as the previous PGT organ pipe 

germanium array system.

ΔB = 0

m = 4

Table B-7 lists the values of variables for the EG&G organ pipe germanium detector system.
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

Table B-7. Values of variables for the EG&G organ 

pipe germanium detector system.

Variable 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-4, 

is given by:

(B-13) 

0 3708 2 0 1 ( )

0 635

42 36e

e

. I.

. I

. K

− −

=

Table B-8 lists calibration factors for the EG&G organ pipe General Electric detector system.

Table B-8. Calibration factors (K) for the sixdetector EG&G organ pipe germanium 

detector system.

Index 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 

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:

(B-14) 2

1 ε a CWT = a

where

ε = counts 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:

(B-15) 0 41 35 9 . CWT K . − =
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

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 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:

(B-16) 

6 1 10

48 2

 241 ppm Am MDA Conversion Factor .

× − =

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 from a possible inhalation 

incident. If a representative sample was not obtained or the origin of the intake was not known, a 

default value of 1,000 ppm 241Am was used and was assigned to the date of the intake or to the date 

of the first positive lung count if the date of the intake was not known. For subsequent lung counts, 

the value of the ppm 241Am was updated to account for the ingrowth of the 241Am from the nuclear 

transformation of 241Pu and for the radioactive decay of the 241Am. The rate of ingrowth of 241Am in 

the plutonium mixture depends on the fraction by weight of the 241Pu in the mixture. The initial weight 

fraction of 241Pu in RFP plutonium was taken to be 0.005 in the 1950s and 1960s and 0.0036 in the 

1970s and later (See Table 5-1). The value of 0.0049 had been historically used at the RFP body 

counter as the rounded value of 0.005. 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.

Table B-10. Americium-241 ingrowth in plutonium.

Initial fraction Pu-241 = 0.0036

Years 100 1,000 10,000

1 270 1,200 10,200

2 430 1,300 10,300

4 730 1,600 10,600

6 1,000 1,900 10,800

10 1,500 2,400 11,200

20 2,300 3,200 11,900

30 2,800 3,600 12,200

40 3,000 3,900 12,300

50 3,200 4,000 12,300
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ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

Initial fraction Pu-241 = 0.0050

Years 100 1,000 10,000

1 340 1,200 10,200

2 560 1,500 10,400

4 980 1,900 10,800

6 1,400 2,200 11,100

10 2,000 2,900 11,700

20 3,100 4,000 12,700

30 3,800 4,700 13,200

40 4,200 5,000 13,500

50 4,400 5,200 13,500

The appropriate value of the ppm 241Am should be applied for lung counts that occurred after a known 

or assumed intake.

The equation to calculate the ppm 241Am for any time (years) after the intake is:

( ) (B-17) Pu241 Am241

Pu239

6

λ λ 0 2

1 0 λ

0 2

1 10

e e

e

T T

T

A L A LP

A L

− −

−

×

= −+

+

where

A = ppm 241Am at time T (yr) 

L1 = λPu241 ÷ (λAm241 – λPu241) 

P0 = initial 241Pu ppm = (initial 241Pu fraction by weight) × (1 × 106

 – A0) 

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

L2 = λAm241

6

0

e

1 10

T

A

−

− × −

λPu239 = decay constant for 239Pu (half-life = 24,100 yr) = 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.
ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 104 of 263

Table B-11. Americium-241 MDA for in vivo lung counts.a,b

Period Detector system Indexc

Minimum system

half timed

Minimum system

full time

Standard system

half timed

Standard system

full time

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

1964–1968e NaI(Tl) 4×4 1.35 2.8 2.5 2.1 1.9

1964–1968e NaI(Tl) 4×4 1.80 4.6 4.1 3.5 3.2

1969–f NaI(Tl) 4×4 0.90 Not applicable Not applicable 0.80 0.76

1969–f NaI(Tl) 4×4 1.35 Not applicable Not applicable 1.3 1.3

1969–f NaI(Tl) 4×4 1.80 Not applicable Not applicable 2.2 2.0

1973–g Phoswich 0.90 Not applicable Not applicable 1.2 1.2

1973–g Phoswich 1.35 Not applicable Not applicable 2.0 2.0

1973–g Phoswich 1.80 Not applicable Not applicable 3.3 3.2

1976–1978h,i Ortec Arrays (high-purity Ge) 0.90 0.26 0.18 0.20 0.14

1976–1978h,i Ortec Arrays (high-purity Ge) 1.35 0.48 0.32 0.37 0.25

1976–1978h,i Ortec Arrays (high-purity Ge) 1.80 0.86 0.59 0.66 0.45

1979–i Ortec Arrays (high-purity Ge) 0.90 0.20 0.14 0.16 0.11

1979–i Ortec Arrays (high-purity Ge) 1.35 0.37 0.25 0.28 0.19

1979–i Ortec Arrays (high-purity Ge) 1.80 0.66 0.45 0.51 0.35

1978–i PGT I Arrays (high-purity Ge) 0.90 0.22 0.15 0.17 0.12

1978–i PGT I Arrays (high-purity Ge) 1.35 0.40 0.27 0.31 0.21

1978–i PGT I Arrays (high-purity Ge) 1.80 0.71 0.49 0.55 0.38

1979–i PGT I Arrays (high-purity Ge) 0.90 0.17 0.12 0.13 0.09

1979–i PGT I Arrays (high-purity Ge) 1.35 0.31 0.21 0.24 0.16

1979–i PGT I Arrays (high-purity Ge) 1.80 0.55 0.38 0.42 0.29

1979–i PGT II Arrays (high-purity Ge) 0.90 0.22 0.15 0.17 0.12

1979–i PGT II Arrays (high-purity Ge) 1.35 0.40 0.28 0.31 0.21

1979–i PGT II Arrays (high-purity Ge) 1.80 0.74 0.50 0.57 0.39

1985–j PGT Organ Pipe (OP) Ge Detectors 0.90 Not applicable Not applicable 0.15 0.11

1985–j PGT Organ Pipe (OP) Ge Detectors 1.35 Not applicable Not applicable 0.26 0.18

1985–j PGT Organ Pipe (OP) Ge Detectors 1.80 Not applicable Not applicable 0.46 0.32
ATTACHMENT B 

MINIMUM DETECTABLE AMOUNTS FOR IN VIVO LUNG COUNTS (continued)

Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 105 of 263

Period Detector system Indexc

Minimum system

half timed

Minimum system

full time

Standard system

half timed

Standard system

full time

1991–k EG&G Organ Pipe Ge Detectors 0.90 Not applicable Not applicable 0.14 0.10

1991–k EG&G Organ Pipe Ge Detectors 1.35 Not applicable Not applicable 0.26 0.18

1991–k EG&G Organ Pipe Ge Detectors 1.80 Not applicable Not applicable 0.48 0.33

1995–l Ortec 2 Organ Pipe Ge Detectors 0.90 Not applicable Not applicable Not applicable 0.14

1995–l Ortec 2 Organ Pipe Ge Detectors 1.35 Not applicable Not applicable Not applicable 0.3

1995–l Ortec 2 Organ Pipe Ge Detectors 1.80 Not applicable Not applicable Not applicable 0.6

a. Americium-241 grows into the plutonium mixture from the nuclear transformation of Pu-241. The initial weight fraction of Pu-241 in RFP plutonium was 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 Am-241, the ppm Am-241 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.

b. To convert the MDA for Am-241 to the MDA for Pu-239,240, multiply the MDA for Am239,240 by [(1 × 106 – ppm Am239,240) ÷ (48.2 × ppm Am239,240)], where ppm 

Am239,240 is the parts per million by weight of the Am239,240 in the plutonium mixture at the time of the lung count.

c. 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).

d. Halved count times were usually used for nonscheduled counts or when a large number of subjects needed to be counted expeditiously.

e, Full time = 40 MLT; minimum system is one detector over the left chest; standard system is two detectors, over right and left chests.

f. Full time = 2,000 s; standard system is two detectors, over right and left chests.

g 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.

h. 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 involved detector types.

i. Full time = 2,000 s; standard system is eight detectors in two array; minimum system is five detectors in two arrays.

j. Standard system = six detectors; standard count time = 2,000 s; occasionally, five or seven detectors were used.

k. Standard system = six detectors; standard count time = 2,000 s.

l. Standard system = four detectors; standard count time = 2,000 s.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 106 of 263

 

ATTACHMENT C 

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS

LIST OF FIGURES

FIGURE TITLE PAGE

C-1 Urinalysis Record Card and HSDS – Urinalysis Detail (1) (first activity date on the HSDS 

portion 10-29-54) ................................................................................................................. 107

C-2 Urinalysis Record Card and HSDS – Urinalysis Detail (2) (first activity date on the HSDS 

portion 8-19-53) ................................................................................................................... 108

C-3 Urinalysis Record Card and HSDS – Urinalysis Detail (3) (first activity date on the HSDS 

portion: 1-6-58) ................................................................................................................... 109

C-4 HSDS – Urinalysis Detail (1) (first activity date 9-17-58) ...................................................... 110

C-5 HSDS – Urinalysis Detail (2) (first activity date 3-19-73) ...................................................... 111

C-6 Analytical Report – Bioassay Analysis Data 3-15-93............................................................ 112

C-7 Analytical Report – Bioassay Analysis Data 10-28-93 .......................................................... 113

C-8 Form 1 – Sample Results 1-29-96........................................................................................ 114

C-9 Rocky Flats Environmental Technology Site (1) 8-27-96...................................................... 115

C-10 Rocky Flats Environmental Technology Site (1) 8-8-96........................................................ 116

C-11 Form 1 – Sample Results – Quanterra, Richland 7-31-98 .................................................... 117

C-12 General Engineering Laboratories 6-28-99 .......................................................................... 118

C-13 Health Sciences Urinalysis Record (with tritium, fecal, and nasal smear results) ................. 119

C-14 Health Physics – Body Counter Information 12-8-65............................................................ 120

C-15 Health Physics – Body Counter Information 5-16-68............................................................ 121

C-16 Health Physics – Body Counter Information 8-26-68............................................................ 122

C-17 Health Physics – Body Counter Information 9-16-70............................................................ 124

C-18 Radiation Dosimetry – Body Count Results 10-3-74............................................................. 124

C-19 Radiation Dosimetry – Body Count Results 5-30-75 ............................................................. 125

C-20 Radiation Dosimetry – Body Count Results 1-9-78............................................................... 127

C-21 Body Counter Results 12-8-81 ............................................................................................. 128

C-22 Radiation Dose Assessment – Body Count Results 7-22-83................................................ 128

C-23 Radiation Dose Assessment – Body Count Results 5-18-83................................................ 130

C-24 Radiation Dose Assessment – Body Count Results 2-21-84................................................ 131

C-25 Radiation Dose Assessment – Body Count Results 3-22-84................................................ 132

C-26 Radiation Dose Assessment – Body Count Results 10-10-85 .............................................. 133

C-27 Radiation Dose Assessment – Body Count Results 3-6-89.................................................. 134

C-28 Internal Dosimetry – Lung Count Results 11-23-93.............................................................. 135

C-29 ABACOS-Plus 3-6-96........................................................................................................... 135

C-30 ABACOS-Plus 11-15-01....................................................................................................... 137

C-31 ABACOS-Plus 6-14-01......................................................................................................... 137
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 107 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-1. Urinalysis Record Card and HSDS – Urinalysis Detail (1) (first activity date on the 

HSDS portion: 10-29-54).
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 108 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-2. Urinalysis Record Card and HSDS – Urinalysis Detail (2) (first activity date on the HSDS 

portion 8-19-53).
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 109 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-3. Urinalysis Record Card and HSDS – Urinalysis Detail (3) (first activity date on the HSDS 

portion: 1-6-58).
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 110 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-4. HSDS – Urinalysis Detail (1) (first activity date 9-17-58).
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 111 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-5. HSDS – Urinalysis Detail (2) (first activity date 3-19-73).
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 112 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-6. Analytical Report – Bioassay Analysis Data 3-15-93.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 113 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-7. Analytical Report – Bioassay Analysis Data 10-28-93.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 114 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-8. Form 1 – Sample Results 1-29-96.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 115 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-9. Rocky Flats Environmental Technology Site (1) 8-27-96 (analytes: 238U, 235U, 234U).
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 116 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-10. Rocky Flats Environmental Technology Site (1) 8-8-96 (analyte: 239Pu).
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 117 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-11. Form 1 – Sample Results – Quanterra, Richland 7-31-98.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 118 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-12. General Engineering Laboratories 6-28-99.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 119 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-13. Health Sciences Urinalysis Record (with tritium, fecal, and nasal smear results).
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 120 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-14. Health Physics – Body Counter Information 12-8-65.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 121 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-15. Health Physics – Body Counter Information 5-16-68.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 122 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-16. Health Physics – Body Counter Information 8-26-68.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 123 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-17. Health Physics – Body Counter Information 9-16-70.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 124 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-18. Radiation Dosimetry – Body Count Results 10-3-74.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 125 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-19. Radiation Dosimetry – Body Count Results 5-30-75.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 126 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-20. Radiation Dosimetry – Body Count Results 1-9-78.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 127 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-21. Body Counter Results 12-8-81.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 128 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-22. Radiation Dose Assessment – Body Count Results 7-22-83.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 129 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-23. Radiation Dose Assessment – Body Count Results 5-18-83.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 130 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-24. Radiation Dose Assessment – Body Count Results 2-21-84.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 131 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-25. Radiation Dose Assessment – Body Count Results 3-22-84.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 132 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-26. Radiation Dose Assessment – Body Count Results 10-10-85.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 133 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-27. Radiation Dose Assessment – Body Count Results 3-6-89.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 134 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-28. Internal Dosimetry – Lung Count Results 11-23-93.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 135 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-29. ABACOS-Plus 3-6-96.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 136 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-30. ABACOS-Plus 11-15-01.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 137 of 263

ATTACHMENT C

EXAMPLES OF BIOASSAY DATA RECORDS AND REPORTS (continued)

Figure C-31. ABACOS-Plus 6-14-01.
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 138 of 263

 

ATTACHMENT D 

INTERNAL CO-EXPOSURE DOSIMETRY DATA

TABLE OF CONTENTS

SECTION TITLE PAGE

D.1 Purpose ............................................................................................................................... 145

D.1.1 Data Overview.......................................................................................................... 145

D.1.2 Bioassay Data Selection........................................................................................... 145

D.1.3 Analysis.................................................................................................................... 146

D.2 Intake Modeling.................................................................................................................... 147

D.2.1 Assumptions............................................................................................................. 147

D.2.2 Bioassay Fitting ........................................................................................................ 147

D.2.3 Material Types.......................................................................................................... 148

D.2.3.1 Uranium ..................................................................................................... 148

D.2.3.2 Plutonium................................................................................................... 148

D.3 Assigning Intakes and Doses ............................................................................................... 149

D.3.1 Intake Rate Summary............................................................................................... 150

D.3.2 Dose Assignment ..................................................................................................... 150

D.4 Co-Exposure Data Tables and Figures ................................................................................ 151

LIST OF TABLES

TABLE TITLE PAGE

D-1 Summary of uranium urinary excretion rate analyses, 1953 to 2005 .................................... 151

D-2 Summary of plutonium urinary excretion rate analyses, 1952 to 2005.................................. 153

D-3 Americium-241 lung count bioassay data............................................................................. 156

D-4 Derived uranium intake rates (95th-percentile), type F material, 1953 to 2005..................... 157

D-5 Derived uranium intake rates (95th-percentile), type M material, 1953 to 2005 .................... 157

D-6 Derived uranium intake rates (95th-percentile), type S material, 1953 to 2005..................... 157

D-7 Derived type M plutonium intake rates (95th-percentile), 1952 to 2005 ................................ 158

D-8 Derived type S 239Pu intake rates (95th-percentile), 1952 to 2005........................................ 158

D-9 Derived type SS 239Pu intake rates (95th-percentile), 1952 to 2005...................................... 158

LIST OF FIGURES

FIGURE TITLE PAGE

D-1 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1976, 50thpercentile, type F ................................................................................................................. 159

D-2 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 50thpercentile, type F ................................................................................................................. 159
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 139 of 263

ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

D-3 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1989 to 12/31/1993, 

50th-percentile, type F ......................................................................................................... 160

D-4 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1994 to 12/31/2005, 

50th-percentile, type F ......................................................................................................... 160

D-5 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/2005, 50th-percentile, type F ...................................................................................... 161

D-6 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1976, 84thpercentile, type F ................................................................................................................. 161

D-7 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 84thpercentile, type F. ................................................................................................................ 162

D-8 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1989 to 12/31/1993, 

84th-percentile, type F ......................................................................................................... 162

D-9 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1994 to 12/31/2005, 

84th-percentile, type F ......................................................................................................... 163

D-10 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/2005, 84th-percentile, type F ...................................................................................... 163

D-11 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1976, 50thpercentile, type M ................................................................................................................. 164

D-12 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 50thpercentile, type M ................................................................................................................. 164

D-13 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1989 to 12/31/1993, 

50th-percentile, type M ......................................................................................................... 165

D-14 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1994 to 12/31/2005, 

50th-percentile, type M ......................................................................................................... 165

D-15 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/2005, 50th-percentile, type M ..................................................................................... 166

D-16 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1988, 

84th-percentile, type M ......................................................................................................... 166

D-17 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1989 to 12/31/1993, 

84th-percentile, type M ......................................................................................................... 167

D-18 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1994 to 12/31/2005, 

84th-percentile, type M ......................................................................................................... 167
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

D-19 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/2005, 84th-percentile, type M ..................................................................................... 168

D-20 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1958, 50thpercentile, type S ................................................................................................................. 168

D-21 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1959 to 12/31/1960, 50thpercentile, type S ................................................................................................................. 169

D-22 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1961 to 12/31/1963, 50thpercentile, type S ................................................................................................................. 169

D-23 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1964 to 12/31/1976, 50thpercentile, type S ................................................................................................................. 170

D-24 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 50thpercentile, type S ................................................................................................................. 170

D-25 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1989 to 12/31/1993, 50thpercentile, type S ................................................................................................................. 171

D-26 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/19e94 to 12/31/2005, 50thpercentile, type S ................................................................................................................. 171

D-27 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/2005, 50th-percentile, type S ...................................................................................... 172

D-28 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1953 to 12/31/1958, 84thpercentile, type S ................................................................................................................. 172

D-29 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1959 to 12/31/1960, 84thpercentile, type S ................................................................................................................. 173

D-30 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1961 to 12/31/1963, 84thpercentile, type S ................................................................................................................. 173

D-31 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1964 to 12/31/1976, 84thpercentile, type S ................................................................................................................. 174

D-32 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1977 to 12/31/1988, 84thpercentile, type S ................................................................................................................. 174

D-33 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1989 to 12/31/1993, 84thpercentile, type S ................................................................................................................. 175

D-34 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results, 1/1/1994 to 12/31/2005, 84thpercentile, type S ................................................................................................................. 175
Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 141 of 263

ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

D-35 Predicted uranium bioassay results calculated using IMBA-derived uranium intake rates 

compared with measured uranium-in-urine results from all intakes 1/1/1953 to 

12/31/2005, 84th-percentile, type S ...................................................................................... 176

D-36 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 50thpercentile, type M ................................................................................................................. 176

D-37 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1969, 50thpercentile, type M ................................................................................................................. 177

D-38 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1970 to 12/31/1979, 50thpercentile, type M ................................................................................................................. 177

D-39 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1980 to 12/31/1988, 50thpercentile, type M ................................................................................................................. 178

D-40 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1989 to 12/31/1993, 50thpercentile, type M................................................................................................................. 178

D-41 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1994 to 12/31/2005, 50thpercentile, type M................................................................................................................. 179

D-42 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, from all intakes 1/1/1952 to 

12/31/1988, 50th-percentile, type M ..................................................................................... 179

D-43 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 84thpercentile, type M ................................................................................................................. 180

D-44 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1969, 84thpercentile, type M ................................................................................................................. 180

D-45 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1970 to 12/31/1979, 84thpercentile, type M ................................................................................................................. 181

D-46 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1980 to 12/31/1988, 84thpercentile, type M ................................................................................................................. 181

D-47 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1989 to 12/31/1993, 84thpercentile, type M ................................................................................................................. 182

D-48 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1994 to 12/31/2005, 84thpercentile, type M ................................................................................................................. 182

D-49 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, from all intakes 1/1/1952 to 

12/31/2005, 84th-percentile, type M ..................................................................................... 183

D-50 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 50thpercentile, type S ................................................................................................................. 183
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

D-51 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1971, 50thpercentile, type S ................................................................................................................. 184

D-52 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1972 to 12/31/1979, 50thpercentile, type S ................................................................................................................. 184

D-53 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1980 to 12/31/1993, 50thpercentile, type S ................................................................................................................. 185

D-54 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1994 to 12/31/2005, 50thpercentile, type S ................................................................................................................. 185

D-55 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/2005, 50thpercentile, type S ................................................................................................................. 186

D-56 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/1961, 84thpercentile, type S ................................................................................................................. 186

D-57 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1962 to 12/31/1971, 84thpercentile, type S ................................................................................................................. 187

D-58 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1972 to 12/31/1979, 84thpercentile, type S ................................................................................................................. 187

D-59 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1980 to 12/31/1993, 84thpercentile, type S ................................................................................................................. 188

D-60 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1994 to 12/31/2005, 84thpercentile, type S ................................................................................................................. 188

D-61 Predicted plutonium bioassay results calculated using IMBA-derived plutonium intake 

rates compared with measured plutonium-in-urine results, 1/1/1952 to 12/31/2005, 84thpercentile, type S ................................................................................................................. 189

D-62 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1972 to 12/31/1976, 

50th-percentile, type S ......................................................................................................... 189

D-63 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1977 to 12/31/1982, 

50th-percentile, type S ......................................................................................................... 190

D-64 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1983 to 12/31/1988, 

50th-percentile, type S ......................................................................................................... 190

D-65 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1972 to 12/31/1988, 

50th-percentile, type S ......................................................................................................... 191

D-66 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1972 to 12/31/1976, 

84th-percentile, type S ......................................................................................................... 191
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INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

D-67 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1977 to 12/31/1982, 

84th-percentile, type S ......................................................................................................... 192

D-68 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1983 to 12/31/1988, 

84th-percentile, type S ......................................................................................................... 192

D-69 Predicted americium bioassay results calculated using IMBA-derived americium intake 

rates compared with measured americium lung burden results, 1/1/1972 to 12/31/1988, 

84th-percentile, type S ......................................................................................................... 193

D-70 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 

12/31/1961, 50th-percentile, type SS ................................................................................... 193

D-71 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1962 to 

12/31/1971, 50th-percentile, type SS ................................................................................... 194

D-72 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1972 to 

12/31/1979, 50th-percentile, type SS ................................................................................... 194

D-73 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 

12/31/1993, 50th-percentile, type SS ................................................................................... 195

D-74 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1994 to 

12/31/2005, 50th-percentile, type SS ................................................................................... 195

D-75 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 

12/31/2005, 50th-percentile, type SS ................................................................................... 196

D-76 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 

12/31/1961, 84th-percentile, type SS ................................................................................... 196

D-77 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1962 to 

12/31/1971, 84th-percentile, type SS ................................................................................... 197

D-78 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1972 to 

12/31/1979, 84th-percentile, type SS ................................................................................... 197

D-79 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 

12/31/1993, 84th-percentile, type SSe ................................................................................. 198

D-80 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1994 to 

12/31/2005, 84th-percentile, type SS ................................................................................... 198

D-81 Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 

12/31/2005, 84th-percentile, type SS ................................................................................... 199

D-82 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1972 to 

12/31/1976, 50th-percentile, type SS ................................................................................... 199
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INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

D-83 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1977 to 

12/31/1982, 50th-percentile, type SS ................................................................................... 200

D-84 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1983 to 

12/31/1988, 50th-percentile, type SS ................................................................................... 200

D-85 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots) from all 

intakes 1/1/1972 to 12/31/1988, 50th-percentile, type SS .................................................... 201

D-86 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1972 to 

12/31/1976, 84th-percentile, type SS ................................................................................... 201

D-87 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1977 to 

12/31/1982, 84th-percentile, type SS ................................................................................... 202

D-88 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1983 to 

12/31/1988, 84th-percentile, type SS ................................................................................... 202

D-89 Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots) from all 

intakes 1/1/1972 to 12/31/1988, 84th-percentile, type SS .................................................... 203
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

D.1 PURPOSE

Some employees at DOE sites were not monitored for internal ionizing radiation exposure, or the 

records of such monitoring are incomplete or unavailable. In such cases, data from monitored 

coworkers can be used to estimate an individual’s possible exposure. The purpose of this attachment 

is to provide monitored co-exposure information for calculating and assigning occupational internal 

doses to employees at RFP for whom no or insufficient bioassay monitoring records exist.

D.1.1 Data Overview

This section provides information on the general selection characteristics of the data and the methods 

of analysis. More detailed radionuclide-specific information is provided in Section D.2. 

D.1.2 Bioassay Data Selection

Urinalysis data used in this assessment for uranium and plutonium from 1952 to 1988 were extracted 

from the CEDR database. There were just over 300,000 records in the urinalysis database. Four 

cases had a date before 1952: one each in 1950 and 1951 and two that appeared to be date errors 

(years incorrectly entered as 1911 and 1923). Urinalysis data used in this assessment for uranium 

and plutonium from 1989 to 2005 were extracted from the HIS-20 database. 

A comparison of CEDR and HIS-20 was made. The databases for the pre-1989 data were

comparable but provide slightly differing results in some cases. These differences sometimes 

suggest CEDR could be slightly more favorable to the claimant while, in other cases, the data suggest 

HIS-20 could be slightly more favorable to the claimant. For the majority of the data, the results are 

similar. In addition, concern was expressed by the Rocky Flats Working Group that the number of 

samples in HIS-20 and CEDR were different in some cases. NIOSH demonstrated that the intakes 

that were predicted by either database were almost identical. However the concern on the part of the 

Working Group persisted. It was suggested and agreed that the use of the 95th percentile internal coexposure intake for unmonitored workers with nontrivial exposure potential would satisfy this concern. 

It was also agreed that this situation and this policy were specific to RFP, and would set no precedent 

to be applied elsewhere.

In most cases, both the uranium and plutonium results were recorded as dpm/24 hr. However, the 

DU units are date-dependent. Through April 1964 the units were μg/24 hr; from May 1964 to 1988 the 

units were dpm/24 hr. Micrograms of uranium were converted to dpm by a 0.89 multiplier determined 

from the IMBA isotopic abundances for DU. Once converted to dpm, the uranium data were assumed 

to be entirely 234

U (Note: This assumption has no impact the statistics). See Section D.3.1 for 

additional discussion on using 234U for the analyses.

All of the uranium and plutonium urinalysis results were recorded either as positive numbers or zeros. 

In general, a zero entry meant the result was less than some reporting level, but actual results were 

reported after April 6, 1970. Zeros were reported in 176,900 records, a little over half of the results for 

all measurements. Uranium and plutonium urinalysis data with a “1” flag in the “nocalc” field of the 

database (about 2,500 records out of roughly 300,000) should be (and were) excluded from analysis 

because the data did not meet quality objectives. For the post-1988 data, values were recorded as 

indicated by the analysis; that is, the actual results of the analyses were recorded even if it was a 

negative number. Because this analysis assumes the data can be represented by a lognormal 

distribution, the data had to be log-transformed. This required the zero and negative results to be 

treated as censored values.
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INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

In vivo 241Am lung data from 1965 to 1988 were extracted from a database table named 

“RFFACW02_RFWB.” There were just fewer than 80,000 241Am records in the lung database. From 

1965 through 1971, all results (about 4,000) were reported as zero, with no explanation of what those 

values might have meant. Therefore, no analyses were performed on those data. The 241Am 

activities were quantified only if a known plutonium incident occurred. However, the results were 

sometimes recorded (in counts per minute) when no known incident had occurred. Some results 

were recorded in micrograms or nanocuries. Therefore, careful interpretation of the data units was 

imperative. Positive values began to appear after 1971, but there still were no exclusion instructions 

for when zero values were reported (see the “nocalc” discussion above). Therefore, zero results were 

treated as zeros because no better information was available. Calculations of the lung plutonium 

values that were recorded with the 241Am lung data were determined by using the 241Am data and an 

assumed concentration of 100 ppm (by weight) of 241Am in the plutonium.

In both the urinalysis and lung-counting datasets, badge numbers (the ID column) are associated with 

most records. However, in the urinalysis data, 55,200 records had a “0” in the badge number column. 

It was not determined what a “0” badge ID meant other than, perhaps, to identify unbadged personnel. 

For the urinalysis data, about 34,000 of the “0” badges were plutonium records; 15,000 were gross 

alpha (A) and 6,000 were uranium (U). It was decided to treat “0” badge numbers as one individual 

when counting the number of unique individuals in any period. The “sdate” column provided the date 

of each analysis in YYMMDD order.

D.1.3 Analysis

Bioassay data were analyzed by quarter or year, depending on the amount of data available during 

the periods. A lognormal distribution was assumed. As mentioned in the previous section, a large 

fraction of the uranium and plutonium urinalysis data were entered as zeros. In many cases, this fact 

made analysis of the data difficult because so few positive values were reported. Therefore, where a 

reporting level was specified and where zeros were inserted for the actual values in the original data 

(below the reporting level), a linear distribution between zero and the reporting level was substituted 

for the zeros. The linear distribution had the form c/n, 2c/n, 3c/n,..., nc/n where n is the number of 

zero values less than the reporting level c. Using R2 as the fit criterion, this linear distribution (alone) 

fits a lognormal transformation by better than 80% and typically significantly improves the goodness of 

fit for the entire dataset. Furthermore, the linear distribution has an average equal to half of the 

reporting value, consistent with the general dose reconstruction practice of assigning half of the lower 

limit of detection for missed dose calculations. As a consequence, substituting a linear distribution for 

these zero entries appears reasonable.

Whenever a linear distribution was substituted for values below a reporting level, the reporting levels 

were used. For EU, these reporting levels were 8.8 dpm/24 hr through 1963, and 20 to 28 dpm/24 hr 

after 1963. For DU, the reporting levels were 5.8 dpm/24 hr through April 1964, 20 to 28 dpm for May 

1964 to 1979, and actual measured values thereafter. For plutonium, these reporting levels were 

0.88 dpm/24 hr through 1961, 0.2 dpm/24 hr for 1962 to April 1970, and actual measured values after 

April 1970. The reporting level for gross alpha through 1963 was 8.8 dpm/24 hr (assigned as EU) and 

0.9 dpm/24 hr thereafter (assigned as plutonium). No reporting level was given for americium-in-lung 

measurements.

After log-transforming the data, the 50th- and 84th-percentile values were determined for each period 

using the method described in ORAUT (2004). Tables D-1, D-2, and D-3 show the statistical analysis 

results for uranium, plutonium, and 241Am, respectively. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

D.2 INTAKE MODELING

This section discusses intake modeling assumptions, intake fitting, and the intake materials (uranium 

and plutonium).

D.2.1 Assumptions

Each result in the intake calculations was assumed to be normally distributed. A uniform absolute 

error of 1 was applied to all results, thus assigning the same weight to each result. IMBA and IDOT 

require results to be in units of activity per day; therefore, all urinalysis results were normalized, as 

needed, to 24-hour samples, using 1,400 mL, the volume of urine excreted by Reference Man in a 

24-hour period.

Because of the nature of work at RFP, it is possible that intakes could have been either chronic or 

acute. However, a series of acute intakes can be approximated as a chronic intake. Therefore, 

intakes were assumed to be chronic and were assumed to occur through inhalation, using a default 

breathing rate of 1.2 m3

/hr and a 5-μm AMAD particle size distribution (ICRP 1995).

For intake modeling purposes, all uranium activity was assumed to be 234U. This assumption does not 

affect the fitting of the data for intake determination because all uranium isotopes behave the same 

biokinetically and the isotopes considered in this analysis all have long half-lives in relation to the 

assumed intake period. ICRP Publication 68 dose coefficients (also referred to as dose conversion 

factors or DCFs) for 234U are 7% to 31% larger than those for 235U, 236U, and 238U (ICRP 1995). 

Therefore, the assumption that the intake is 100% 234U provides a result favorable to the claimant.

For plutonium, 239Pu was assumed for the intake modeling. Before the mid-1970s, plutonium 

urinalysis was performed by chemical separation followed by the counting of all alpha-emitting 

isotopes of plutonium (i.e., 238Pu, 239Pu, and 240Pu). In the mid-1970s, alpha spectroscopy was used 

to differentiate between them. For this modeling, the gross plutonium alpha results are assumed to 

represent only alphas from 239Pu, which results in approximately a 2% overestimate of the modeled 

intakes. This assumption is made to enable consistent modeling of data from both types of urinalysis.

Starting in 1972, lung counts were performed to determine the lung burden of 241Am. These lung 

counts can be used to determine the intake of plutonium. For each plutonium material type, the more 

limiting value of the intakes as determined by the americium lung counts or plutonium urinalysis was 

used. Use of the higher value (from the less sensitive bioassay method for a given material type) 

would be inconsistent with the available bioassay records because a higher intake would result in 

higher-than-observed bioassay results from the more sensitive bioassay method.

D.2.2 Bioassay Fitting

IMBA and IDOT were used to fit the bioassay results to a series of inhalation intakes. Data from 1952 

through 2005 were fit as a series of chronic intakes. The intake assumptions were based on observed 

patterns in the bioassay data. Periods with constant chronic intake rates were chosen by selecting 

periods where the bioassay results were similar. A new chronic intake period was started if the data 

indicated a significant sustained change in the bioassay results. By this method, 1952 through 2005

were divided into multiple chronic intake periods.
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INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

D.2.3 Material Types

See Section 5.2 for source term solubilities.

D.2.3.1 Uranium

Because the uranium isotopes at RFP have very long radiological half-lives and the material is 

retained in the body for long periods, excretion results are not independent. For example, an intake in 

the 1950s could contribute to urinary excretion in the 1980s and later. To avoid potential 

underestimation of intakes for people who worked at RFP for relatively short periods, each chronic 

intake was fit independently, using only the bioassay results from the single intake period for type S 

solubility. For type M and F solubilities, the approach used determined that earlier intake rates 

significantly biased later intake rates. This method results in a potential overestimate of intakes for 

exposures extending through multiple assumed intake periods. Only the results within the intake 

period were selected for use in fitting each period. Excluded results are shown in light gray in the 

figures at the end of this attachment.

Uranium urinalysis results were analyzed with IMBA to derive intake rates for 1953 to 2005. Excretion 

data are shown in Table D-1. 

Uranium Type F

The solid lines in Figures D-1 through D-4 show the individual fits to the 50th-percentile excretion 

rates for type F material. Figure D-5 is the combined fit for all the intake periods. Figures D-6 through 

D-9 show the individual fits to the 84th-percentile excretion rates for type F uranium material. Figure 

D-10 shows the overall fit to the 84th-percentile excretion rates for type F uranium material. Table D4 tabulates the derived intake rates at both the 50th- and 84th-percentile levels along with the 

associated geometric standard deviations (GSDs). 

Uranium Type M

Figures D-11 through D-14 show the individual fits to the 50th-percentile excretion rates for type M 

uranium material. Figure D-15 is the combined fit for all the intake periods. Figures D-16 through D18 show the individual fits to the 84th-percentile excretion rates for type M uranium material. Figure 

D-19 shows the overall fit to the 84th-percentile excretion rates for type M uranium material. Table D5 tabulates the derived intake rates at both the 50th- and 84th-percentile levels along with the 

associated GSDs.

Uranium Type S

Figures D-20 through D-26 show the individual fits to the 50th-percentile excretion rates for type S 

uranium material. Figure D-27 is the combined fit for all the intake periods. Figures D-28 through D34 show the individual fits to the 84th-percentile excretion rates for type S uranium material. Figure 

D-35 shows the overall fit to the 84th-percentile excretion rates for type S uranium material. Table D6 tabulates the derived intake rates at both the 50th- and 84th-percentile levels along with the 

associated GSDs. 

D.2.3.2 Plutonium

Plutonium urinalysis results were analyzed with IMBA using type M and S materials to derive intake 

rates for 1952 to 2005 and IDOT using type SS material to derive intake rates for 1952 to 2005. As 

with type S uranium, plutonium isotopes at RFP have very long radiological half-lives and the material 

is retained in the body for long periods, so excretion results are not independent. To avoid potential 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

underestimation of intakes for people who worked at RFP for relatively short periods, each chronic 

intake was fit independently, using only the bioassay results from the single intake period for type M, 

S and SS solubility. This method results in a potential overestimate of intakes for exposures 

extending through multiple assumed intake periods. Only the results within the intake period were 

selected for use in fitting each period. Excluded results are shown in light gray in the figures. Tables 

D-2 and D-3 provide the bioassay data that were used to perform the fits.

Plutonium Type M

The solid lines in Figures D-36 through D-41 show the individual fits to the 50th-percentile excretion 

rates for type M plutonium material. Figure D-42 is the combined fit for all the intake periods. Figures 

D-43 through D-48 show the individual fits to the 84th-percentile excretion rates for type M plutonium 

material. Figure D-49 is the combined fit for all the intake periods. Table D-7 lists the 50th- and 84thpercentile intake rates along with the associated GSDs determined from the plutonium urinalysis. 

Plutonium Type S

The solid lines in Figures D-50 through D-54 show the individual fits to the 50th-percentile excretion 

rates for type S plutonium material. Figure D-55 is the combined fit for all the intake periods. Figures 

D-56 through D-60 show the individual fits to the 84th-percentile excretion rates for type S plutonium 

material. Figure D-61 is the combined fit for all the intake periods.

The solid lines in Figures D-62 through D-64 show the individual fits to the 50th-percentile 241Am lung 

burdens for type S material. Figure D-65 is the combined fit for all the intake periods. Figures D-66 

through D-68 show the individual fits to the 84th-percentile 241Am lung burdens for type S material. 

Figure D-69 is the combined fit for all the intake periods.

Table D-8 lists the 50th- and 84th-percentile intake rates along with the associated GSD determined 

from the plutonium urinalysis and 241Am lung count data. 

Plutonium Type SS

The solid lines in Figures D-70 through D-74 show the individual fits to the 50th-percentile excretion 

rates for type SS plutonium material. Figure D-75 is the combined fit for all the intake periods. 

Figures D-76 through D-80 show the individual fits to the 84th-percentile excretion rates for type SS 

plutonium material. Figure D-81 is the combined fit for all the intake periods.

The solid lines in Figures D-82 through D-84 show the individual fits to the 50th-percentile 241Am lung 

burdens for type SS material. Figure D-85 is the combined fit for all the intake periods. Figures D-86 

through D-88 show the individual fits to the 84th-percentile 241Am lung burdens for type SS material. 

Figure D-89 is the combined fit for all the intake periods.

Table D-9 lists the 50th- and 84th-percentile intake rates along with the associated GSD determined 

from the plutonium urinalysis and 241Am lung count data. 

D.3 ASSIGNING INTAKES AND DOSES

This section describes the derived intake rates and provides guidance for assigning doses. For each 

intake period discussed below, the 50th- and 84th-percentile calculated intakes were used to 

determine the GSD of the data. The GSD along with the geometric mean were used to calculate the 

95th-percentile intake rate. In 1993, the Secretary of Energy formally announced the end of nuclear 

production at RFP. Remediation was completed at the RFP in late 2005. Co-exposure intakes 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

should be assigned, when applicable, up through 2005. Only environmental intakes should be 

assigned after 2005.

D.3.1 Intake Rate Summary

Multiple intake periods were fit to the derived 50th- and 84th-percentile uranium excretion data. 

Tables D-4 through D-6 summarize the 95th-percentile uranium intake rates derived from the fits. 

Similarly, multiple intake periods were fit to the derived 50th- and 84th-percentile plutonium excretion 

and americium lung burden data for type M, S, and SS material. Tables D-7 through D-9 summarize 

the 95th-percentile plutonium intake rates derived from the fits.

D.3.2 Dose Assignment

Doses to be assigned to individuals are calculated from the 95th-percentile intake rates. Dose 

reconstructors should select the material type that is the most favorable to the claimant. The constant 

distribution is selected in IREP, with the calculated dose entered as Parameter 1.
ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 151 of 263

D.4 CO-EXPOSURE DATA TABLES AND FIGURES

Table D-1. Summary of uranium urinary excretion rate analyses, 1953 to 2005. 

Effective sample 

date

50th percentile 

(dpm/24 hr)

84th percentile 

(dpm/24 hr)

7/1/1953 3.727E+00 1.001E+01

2/15/1954 3.866E+00 1.036E+01

5/15/1954 4.161E+00 1.147E+01

8/15/1954 3.732E+00 1.007E+01

11/15/1954 3.409E+00 9.389E+00

2/15/1955 3.225E+00 9.019E+00

5/15/1955 3.333E+00 9.487E+00

8/15/1955 3.434E+00 9.406E+00

11/15/1955 3.442E+00 9.875E+00

2/15/1956 3.310E+00 9.039E+00

5/15/1956 3.497E+00 9.843E+00

8/15/1956 3.635E+00 1.021E+01

11/15/1956 3.302E+00 9.121E+00

2/15/1957 3.460E+00 9.894E+00

5/15/1957 3.492E+00 1.017E+01

8/15/1957 3.655E+00 1.078E+01

11/15/1957 3.700E+00 1.100E+01

2/15/1958 4.089E+00 1.258E+01

5/15/1958 3.739E+00 1.059E+01

8/15/1958 3.907E+00 1.127E+01

11/15/1958 4.705E+00 1.432E+01

2/15/1959 4.381E+00 1.316E+01

5/15/1959 5.518E+00 1.791E+01

8/15/1959 5.544E+00 1.657E+01

11/15/1959 5.887E+00 1.913E+01

2/15/1960 8.806E+00 3.307E+01

5/15/1960 6.856E+00 2.223E+01

8/15/1960 7.476E+00 2.421E+01

11/15/1960 6.602E+00 2.367E+01

2/15/1961 5.944E+00 2.026E+01

5/15/1961 5.722E+00 1.863E+01

8/15/1961 5.574E+00 1.829E+01

11/15/1961 6.598E+00 2.267E+01

Effective sample 

date

50th percentile 

(dpm/24 hr)

84th percentile 

(dpm/24 hr)

2/15/1962 5.862E+00 2.045E+01

5/15/1962 4.692E+00 1.538E+01

8/15/1962 5.654E+00 1.674E+01

11/15/1962 4.397E+00 1.383E+01

2/15/1963 4.166E+00 1.323E+01

5/15/1963 4.175E+00 1.315E+01

8/15/1963 3.841E+00 1.228E+01

11/15/1963 3.601E+00 1.151E+01

2/15/1964 6.354E+00 1.851E+01

5/15/1964 8.368E+00 2.339E+01

8/15/1964 8.161E+00 2.217E+01

11/15/1964 8.297E+00 2.354E+01

7/1/1965 7.823E+00 2.079E+01

7/1/1966 7.432E+00 1.836E+01

7/1/1967 7.445E+00 1.844E+01

7/1/1968 7.430E+00 1.846E+01

7/1/1969 7.509E+00 1.852E+01

7/1/1970 7.440E+00 1.828E+01

7/1/1971 7.421E+00 1.813E+01

7/1/1972 7.316E+00 1.818E+01

7/1/1973 7.403E+00 1.806E+01

7/1/1974 7.388E+00 1.808E+01

7/1/1975 7.378E+00 1.810E+01

7/1/1976 7.418E+00 1.804E+01

7/1/1977 1.720E-01 5.380E-01

7/1/1978 8.930E-01 2.355E+00

7/1/1979 4.440E-01 2.037E+00

7/1/1980 2.410E-01 1.049E+00

7/1/1981 1.780E-01 1.109E+00

2/15/1982 2.370E-01 1.152E+00

5/15/1982 6.200E-02 6.770E-01

8/15/1982 1.600E-02 2.110E-01

11/15/1982 1.120E-01 7.410E-01
ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 152 of 263

Effective sample 

date

50th percentile 

(dpm/24 hr)

84th percentile 

(dpm/24 hr)

2/15/1983 2.210E-01 1.062E+00

5/15/1983 4.320E-01 1.330E+00

8/15/1983 3.270E-01 1.576E+00

11/15/1983 7.200E-02 6.460E-01

2/15/1984 2.730E-01 1.400E+00

5/15/1984 2.210E-01 1.330E+00

8/15/1984 1.330E-01 9.970E-01

11/15/1984 6.500E-02 4.640E-01

2/15/1985 3.400E-02 4.100E-01

5/15/1985 3.000E-02 2.810E-01

8/15/1985 4.000E-02 5.110E-01

11/15/1985 3.700E-02 4.150E-01

2/15/1986 2.900E-02 3.570E-01

5/15/1986 3.300E-02 3.390E-01

8/15/1986 1.800E-02 2.070E-01

11/15/1986 2.200E-02 3.160E-01

7/1/1987 5.700E-02 4.670E-01

7/1/1988 5.900E-02 4.120E-01

2/15/1989 5.400E-02 7.260E-01

5/15/1989 4.200E-02 3.370E-01

8/15/1989 3.500E-02 2.580E-01

11/15/1989 5.300E-02 2.900E-01

7/1/1990 3.300E-02 2.120E-01

7/1/1991 3.700E-02 1.870E-01

7/1/1992 6.700E-02 6.780E-01

7/1/1993 4.800E-02 2.780E-01

2/15/1994 8.000E-03 7.100E-02

5/15/1994 2.700E-02 1.820E-01

8/15/1994 1.000E-02 1.000E-01

11/15/1994 2.100E-02 1.260E-01

7/1/1995 2.300E-02 1.240E-01

2/15/1996 6.000E-03 7.500E-02

5/15/1996 3.000E-03 4.700E-02

8/15/1996 4.000E-03 4.200E-02

11/15/1996 7.000E-03 4.700E-02

2/15/1997 1.000E-02 6.700E-02

Effective sample 

date

50th percentile 

(dpm/24 hr)

84th percentile 

(dpm/24 hr)

5/15/1997 1.400E-02 7.600E-02

8/15/1997 1.400E-02 6.000E-02

11/15/1997 1.700E-02 6.000E-02

7/1/1998 4.000E-03 2.400E-02

7/1/1999 8.000E-03 4.600E-02

2/15/2000 7.000E-03 4.600E-02

5/15/2000 4.000E-03 8.900E-02

8/15/2000 4.000E-03 3.600E-02

11/15/2000 5.000E-03 3.800E-02

2/15/2001 1.000E-02 4.100E-02

5/15/2001 1.100E-02 6.600E-02

8/15/2001 7.000E-03 4.900E-02

11/15/2001 6.000E-03 6.800E-02

2/15/2002 9.000E-03 7.200E-02

5/15/2002 1.600E-02 1.420E-01

8/15/2002 1.200E-02 6.550E-01

11/15/2002 9.000E-03 1.690E-01

2/15/2003 6.000E-03 1.100E-01

5/15/2003 1.200E-02 7.400E-02

8/15/2003 3.000E-03 3.200E-02

11/15/2003 1.400E-02 5.400E-02

2/15/2004 1.800E-02 8.900E-02

5/15/2004 1.500E-02 7.900E-02

8/15/2004 8.000E-03 5.500E-02

11/15/2004 7.000E-03 4.940E-01

7/1/2005 1.200E-02 4.800E-02
ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 153 of 263

Table D-2. Summary of plutonium urinary excretion rate analyses, 1952 to 2005.

a,b

Effective

sample date

50th percentile 

(dpm/24 hr)

84th percentile 

(dpm/24 hr)

7/1/1952 2.514E+00 8.198E+00

7/1/1953 7.160E-01 1.046E+00

7/1/1954 5.750E-01 1.053E+00

7/1/1955 4.690E-01 9.190E-01

7/1/1956 6.150E-01 1.264E+00

7/1/1957 2.610E+00 1.201E+01

2/15/1958 2.173E+00 1.004E+01

5/15/1958 1.037E+00 2.872E+00

8/15/1958 1.295E+00 3.801E+00

11/15/1958 9.190E-01 2.581E+00

2/15/1959 7.090E-01 1.542E+00

5/15/1959 9.420E-01 2.276E+00

8/15/1959 9.450E-01 2.482E+00

11/15/1959 5.600E-01 1.211E+00

2/15/1960 6.140E-01 1.353E+00

5/15/1960 5.960E-01 1.221E+00

8/15/1960 4.530E-01 9.550E-01

11/15/1960 5.730E-01 1.528E+00

2/15/1961 7.280E-01 1.625E+00

5/15/1961 6.910E-01 1.377E+00

8/15/1961 7.540E-01 2.035E+00

11/15/1961 6.560E-01 1.645E+00

2/15/1962 3.370E-01 8.090E-01

5/15/1962 3.260E-01 7.350E-01

8/15/1962 2.710E-01 5.890E-01

11/15/1962 2.200E-01 4.310E-01

2/15/1963 2.500E-01 4.670E-01

5/15/1963 2.480E-01 4.960E-01

8/15/1963 2.380E-01 4.320E-01

11/15/1963 2.520E-01 5.620E-01

2/15/1964 2.960E-01 8.100E-01

5/15/1964 2.490E-01 4.830E-01

8/15/1964 3.790E-01 1.668E+00

11/15/1964 3.340E-01 1.066E+00

2/15/1965 2.830E-01 7.570E-01

5/15/1965 3.480E-01 1.085E+00

Effective

sample date

50th percentile 

(dpm/24 hr)

84th percentile 

(dpm/24 hr)

8/15/1965 2.210E-01 4.170E-01

11/15/1965 2.660E-01 6.460E-01

2/15/1966 2.930E-01 8.210E-01

5/15/1966 2.370E-01 5.540E-01

8/15/1966 2.130E-01 4.300E-01

11/15/1966 2.520E-01 6.250E-01

2/15/1967 2.510E-01 6.220E-01

5/15/1967 2.400E-01 5.650E-01

8/15/1967 1.990E-01 4.130E-01

11/15/1967 2.360E-01 5.350E-01

2/15/1968 2.280E-01 5.260E-01

5/15/1968 2.050E-01 4.610E-01

8/15/1968 2.520E-01 5.850E-01

11/15/1968 2.780E-01 7.240E-01

2/15/1969 2.920E-01 6.920E-01

5/15/1969 2.660E-01 6.060E-01

8/15/1969 2.400E-01 5.190E-01

11/15/1969 2.640E-01 5.580E-01

2/15/1970 2.420E-01 5.150E-01

5/15/1970 1.650E-01 6.230E-01

8/15/1970 1.000E-01 4.230E-01

11/15/1970 1.200E-01 4.700E-01

2/15/1971 9.100E-02 3.660E-01

5/15/1971 5.500E-02 2.090E-01

8/15/1971 7.300E-02 2.930E-01

11/15/1971 6.100E-02 2.490E-01

2/15/1972 4.600E-02 3.980E-01

5/15/1972 4.600E-02 4.420E-01

8/15/1972 2.900E-02 1.990E-01

11/15/1972 2.800E-02 1.680E-01

2/15/1973 2.400E-02 1.450E-01

5/15/1973 3.300E-02 1.800E-01

8/15/1973 6.700E-02 3.050E-01

11/15/1973 6.100E-02 2.680E-01

2/15/1974 6.000E-02 2.240E-01

5/15/1974 4.900E-02 1.890E-01
ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 154 of 263

Effective

sample date

50th percentile 

(dpm/24 hr)

84th percentile 

(dpm/24 hr)

8/15/1974 3.300E-02 1.440E-01

11/15/1974 1.600E-02 1.090E-01

2/15/1975 2.100E-02 1.040E-01

5/15/1975 1.900E-02 9.500E-02

8/15/1975 2.200E-02 2.000E-01

11/15/1975 1.500E-02 9.700E-02

2/15/1976 1.600E-02 1.440E-01

5/15/1976 2.100E-02 1.020E-01

8/15/1976 1.500E-02 1.040E-01

11/15/1976 4.300E-02 1.840E-01

2/15/1977 8.300E-02 2.620E-01

5/15/1977 9.200E-02 2.450E-01

8/15/1977 7.200E-02 1.900E-01

11/15/1977 6.200E-02 1.880E-01

2/15/1978 9.500E-02 3.070E-01

5/15/1978 6.000E-02 1.990E-01

8/15/1978 5.600E-02 2.010E-01

11/15/1978 3.300E-02 1.340E-01

2/15/1979 6.200E-02 2.370E-01

5/15/1979 1.300E-02 1.000E-01

8/15/1979 1.300E-02 8.700E-02

11/15/1979 2.900E-02 1.390E-01

2/15/1980 1.700E-02 1.060E-01

5/15/1980 1.700E-02 6.400E-02

8/15/1980 1.300E-02 6.100E-02

11/15/1980 4.000E-03 3.500E-02

2/15/1981 6.000E-03 3.700E-02

8/15/1981 5.000E-03 3.600E-02

11/15/1981 8.000E-03 5.600E-02

2/15/1983 1.000E-03 1.700E-02

8/15/1983 2.000E-03 1.600E-02

11/15/1983 4.000E-03 2.900E-02

2/15/1984 8.000E-03 5.000E-02

5/15/1984 5.300E-02 2.220E-01

8/15/1984 1.100E-02 7.100E-02

11/15/1984 5.400E-02 1.960E-01

2/15/1985 1.000E-02 8.000E-02

Effective

sample date

50th percentile 

(dpm/24 hr)

84th percentile 

(dpm/24 hr)

5/15/1985 2.500E-02 1.000E-01

8/15/1985 1.400E-02 8.100E-02

11/15/1985 1.700E-02 1.000E-01

2/15/1986 5.000E-03 3.300E-02

5/15/1986 4.000E-03 3.800E-02

8/15/1986 7.000E-03 3.800E-02

11/15/1986 8.000E-03 4.200E-02

2/15/1987 4.000E-03 3.000E-02

5/15/1987 5.000E-03 3.600E-02

8/15/1987 8.000E-03 5.100E-02

11/15/1987 8.000E-03 5.000E-02

2/15/1988 3.000E-03 3.200E-02

5/15/1988 2.000E-03 3.300E-02

8/15/1988 5.000E-03 3.400E-02

11/15/1988 6.000E-03 3.800E-02

2/15/1989 4.230E-03 3.460E-02

5/15/1989 1.060E-02 7.150E-02

8/15/1989 2.880E-03 6.550E-02

11/15/1989 6.970E-03 8.480E-02

2/15/1990 2.280E-03 2.830E-02

5/15/1990 3.970E-03 3.090E-02

8/15/1990 2.300E-03 3.070E-02

11/15/1990 3.970E-03 3.760E-02

2/15/1991 4.120E-03 2.560E-02

5/15/1991 1.180E-03 6.270E-02

8/15/1991 7.960E-03 9.720E-02

11/15/1991 6.950E-04 3.900E-02

2/15/1992 5.110E-04 1.910E-02

5/15/1992 2.780E-03 2.020E-02

8/15/1992 5.150E-03 3.630E-02

11/15/1992 9.490E-03 1.070E-01

7/1/1993 4.130E-03 6.460E-02

2/15/1994 4.500E-04 5.460E-03

5/15/1994 5.760E-04 5.650E-03

8/15/1994 6.550E-04 7.430E-03

11/15/1994 8.510E-04 6.880E-03

7/1/1995 8.570E-04 8.150E-03
ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 155 of 263

Effective

sample date

50th percentile 

(dpm/24 hr)

84th percentile 

(dpm/24 hr)

2/15/1996 1.050E-03 9.700E-03

5/15/1996 8.740E-04 7.070E-03

8/15/1996 4.900E-04 4.330E-03

11/15/1996 3.640E-04 5.270E-03

2/15/1997 6.430E-04 7.300E-03

5/15/1997 5.280E-04 6.210E-03

8/15/1997 4.840E-04 6.610E-03

11/15/1997 2.300E-04 1.210E-03

7/1/1998 3.940E-04 3.180E-03

7/1/1999 3.580E-04 4.280E-03

2/15/2000 3.320E-04 2.790E-03

5/15/2000 5.430E-04 3.320E-03

8/15/2000 5.230E-04 3.410E-03

11/15/2000 4.300E-04 2.950E-03

2/15/2001 3.250E-04 3.030E-03

5/15/2001 2.520E-04 4.150E-03

Effective

sample date

50th percentile 

(dpm/24 hr)

84th percentile 

(dpm/24 hr)

8/15/2001 3.940E-04 4.340E-03

11/15/2001 4.190E-04 3.500E-03

2/15/2002 3.860E-04 3.560E-03

5/15/2002 4.050E-04 3.570E-03

8/15/2002 2.420E-04 2.700E-03

11/15/2002 3.630E-04 2.950E-03

2/15/2003 2.320E-04 2.140E-03

5/15/2003 4.720E-04 2.970E-03

8/15/2003 3.150E-04 2.630E-03

11/15/2003 3.050E-04 2.620E-03

2/15/2004 2.840E-04 6.720E-03

5/15/2004 2.690E-04 2.550E-03

8/15/2004 3.620E-04 2.680E-03

11/15/2004 1.790E-04 2.000E-03

7/1/2005 2.040E-04 3.090E-03

a. Very large results for badge 395943 excluded from 1964-1965; badges 164455 and 184168 excluded from quarter 3, 1971; 164455 and 184169 excluded from 

quarter 4, 1971; badge 184106 excluded from quarter 2, 1976.

b. Results for quarter 2, 1981, all of 1982, and quarter 2, 1983 were not used in calculations because there are too few results. 
ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 156 of 263

Table D-3. Americium-241 lung count bioassay data.

a 

Effective sample 

date

50th percentile 

(dpm)

84th percentile 

(dpm)

7/1/1972 0.15 6.66

2/15/1973 11.10 130.98

5/15/1973 22.20 237.54

8/15/1973 55.50 417.36

11/15/1973 19.98 210.90

2/15/1974 11.10 148.74

5/15/1974 15.54 177.60

8/15/1974 15.54 175.38

11/15/1974 15.54 175.38

2/15/1975 38.41 333.00

5/15/1975 59.94 399.60

8/15/1975 87.02 541.68

11/15/1975 105.89 617.16

2/15/1976 95.46 579.42

5/15/1976 97.68 563.88

8/15/1976 37.74 295.26

11/15/1976 26.64 246.42

2/15/1977 22.20 215.34

5/15/1977 17.76 182.04

8/15/1977 15.54 135.42

11/15/1977 8.88 113.22

2/15/1978 17.76 184.26

5/15/1978 15.54 155.40

8/15/1978 15.54 146.52

11/15/1978 8.88 99.90

7/1/1979 26.64 239.76

2/15/1980 57.72 432.90

5/15/1980 44.40 352.98

8/15/1980 46.62 379.62

11/15/1980 59.94 459.54

2/15/1981 39.96 335.22

5/15/1981 35.52 310.80

Effective sample 

date

50th percentile 

(dpm)

84th percentile 

(dpm)

8/15/1981 35.52 306.36

11/15/1981 35.52 301.92

2/15/1982 28.86 279.72

5/15/1982 24.42 246.42

8/15/1982 22.20 226.44

11/15/1982 19.98 179.82

2/15/1983 13.32 146.52

5/15/1983 4.44 68.82

8/15/1983 11.10 122.10

11/15/1983 17.76 139.86

2/15/1984 11.10 128.76

5/15/1984 13.32 128.76

8/15/1984 11.10 119.88

11/15/1984 17.76 148.74

2/15/1985 8.88 93.24

5/15/1985 11.10 113.22

8/15/1985 6.66 77.70

11/15/1985 6.66 82.14

2/15/1986 8.88 108.78

5/15/1986 15.54 119.88

8/15/1986 11.10 126.54

11/15/1986 8.88 95.46

2/15/1987 17.76 159.84

5/15/1987 11.10 113.22

8/15/1987 19.98 202.02

11/15/1987 19.98 159.84

2/15/1988 13.32 135.42

5/15/1988 17.76 162.06

8/15/1988 11.10 95.46

11/15/1988 8.88 93.24

2/15/1989 31.08 208.68
ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Document No. ORAUT-TKBS-0011-5 Revision No. 04 Effective Date: 09/01/2020 Page 157 of 263

Table D-4. Derived uranium intake rates (95th-percentile), dpm/day, type F material, 1953 to 2005.

Period 50th percentile 84th percentile GSD 95th percentile

1953–1958 13.37 37.91 2.84 74.2

1959 19.7 61.99 3.15 130

1960 27.23 94.74 3.48 212

1961 21.62 65.97 3.05 135

1962 16.27 65.97 4.05 163

1963 16.27 44.36 2.73 84.7

1964 27.26 80.39 2.95 161

1965–1976 27.26 66.26 2.43 118

1977–1988 0.597 3.003 5.03 8.52

1989–1993 0.171 1.365 7.98 5.21

1994–2005 0.038 0.370 9.86 1.64

Table D-5. Derived uranium intake rates (95th-percentile), dpm/day, type M material, 1953 to 2005.

Period 50th percentile 84th percentile GSD 95th percentile

1953–1958 54.75 154.8 2.83 303

1959–1960 102.7 347.6 3.38 763

1961–1963 71.85 234.2 3.26 502

1964–1976 112.8 284.2 2.52 516

1977–1988 2.443 6.252 2.56 11.5

1989–1993 0.753 5.821 7.73 21.8

1994–2005 0.154 1.532 9.94 6.74

Table D-6. Derived uranium intake rates (95th-percentile), dpm/day, type S material, 1953 to 2005.

Period 50th percentile 84th percentile GSD 95th percentile

1953–1958 936.9 2,676 2.86 5,266

1959–1960 2,768 9,300 3.36 20,322

1961–1963 1,680 5,439 3.24 11,604

1964–1976 1,630 4,086 2.51 7,391

1977–1988 29.12 157 5.39 465

1989–1993 15.11 115.00 7.61 426

1994–2005 2.06 21.96 10.65 101
ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

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Table D-7. Derived type M plutonium intake rates (95th-percentile), dpm/day, 1952 to 2005. 

Period 50th percentile 84th percentile GSD 95th percentile

1952–1961 121 357.2 2.95 718

1962–1969 43.5 106.5 2.45 190

1970–1979 7.05 29.81 4.23 75.6

1980–1988 1.622 8.908 5.49 26.7

1989–1993 0.9542 11.66 12.22 58.6

1994–2005 0.05418 0.526 9.71 2.28

Table D-8. Derived type S 239Pu intake rates (95th percentile), dpm/day, 1952 to 2005. 

Period 50th percentile 84th percentile GSD 95th percentile 

1952–1961 1,925 5,628 2.92 11,243

1962–1971 518.3 1,362 2.63 2,540

1972–1976 20.59 144.2 7.00 506

1977–1982 10.88 94.56 8.69 381

1983–1988 4.531 43.85 9.68 190

1989–1993 13.7 104.0 7.59 384

1994–2005 0.7966 7.745 9.72 33.6

Table D-9. Derived type SS 239Pu intake rates (95th percentile), dpm/day, 1952 to 2005. 

Period 50th percentile 84th percentile GSD 95th percentilea

1952–1961 16,400 48,500 2.96 99,933

1962–1971 4,470 11,600 2.60 27,238

1972–1976 67.2 465 6.92 1,619

1977–1982 29.9 261 8.74 1,057

1983–1988 12.8 123 9.60 527

1989–1993 130 933 7.18 3326

1994–2005 7.32 71.1 9.71 308

a. A minimum GSD of 3.00 was used to calculated the 95th percentile.
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-1. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1953 to 12/31/1976, 

50th-percentile, type F. 

Figure D-2. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1977 to 12/31/1988, 

50th-percentile, type F.
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-3. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1989 to 12/31/1993, 

50th-percentile, type F.

Figure D-4. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1994 to 12/31/2005, 

50th-percentile, type F. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-5. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots) from all intakes 1/1/1953 to 

12/31/2005, 50th-percentile, type F. 

Figure D-6. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1953 to 12/31/1976, 

84th-percentile, type F. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-7. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1977 to 12/31/1988, 

84th-percentile, type F.

Figure D-8. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1989 to 12/31/1993, 

84th-percentile, type F. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-9. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1994 to 12/31/2005, 

84th-percentile, type F. 

Figure D-10. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots) from all intakes 1/1/1953 to 

12/31/2005, 84th-percentile, type F. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

Figure D-11. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1953 to 12/31/1976, 

50th-percentile, type M. 

Figure D-12. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1977 to 12/31/1988, 

50th-percentile, type M. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-13. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1989 to 12/31/1993, 

50th-percentile, type M. 

Figure D-14. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1994 to 12/31/2005, 

50th-percentile, type M. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-15. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots) from all intakes 1/1/1953 to 

12/31/2005, 50th-percentile, type M. 

Figure D-16. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1953 to 12/31/1988, 

84th-percentile, type M. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-17. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1989 to 12/31/1993, 

84th-percentile, type M. 

Figure D-18. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1994 to 12/31/2005, 

84th-percentile, type M. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-19. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots) from all intakes 1/1/1953 to 

12/31/2005, 84th-percentile, type M. 

Figure D-20. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1953 to 12/31/1958, 

50th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-21. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1959 to 12/31/1960, 

50th-percentile, type S. 

Figure D-22. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1961 to 12/31/1963, 

50th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-23. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1964 to 12/31/1976, 

50th-percentile, type S. 

Figure D-24. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1977 to 12/31/1988, 

50th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-25. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1989 to 12/31/1993, 

50th-percentile, type S. 

Figure D-26. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1994 to 12/31/2005, 

50th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-27. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots) from all intakes 1/1/1953 to 

12/31/2005, 50th-percentile, type S. 

Figure D-28. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1953 to 12/31/1958, 

84th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-29. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1959 to 12/31/1960, 

84th-percentile, type S. 

Figure D-30. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1961 to 12/31/1963, 

84th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued)

Figure D-31. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1964 to 12/31/1976, 

84th-percentile, type S. 

Figure D-12. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1977 to 12/31/1988, 

50th-percentile, type M. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-33. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1989 to 12/31/1993, 

84th-percentile, type S. 

Figure D-34. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots), 1/1/1994 to 12/31/2005, 

84th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-35. Predicted uranium bioassay results (line) calculated using IMBA-derived uranium intake 

rates compared with measured uranium-in-urine results (dots) from all intakes 1/1/1953 to 

12/31/2005, 84th-percentile, type S. 

Figure D-36. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/1961, 

50th-percentile, type M. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-37. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1962 to 12/31/1969, 

50th-percentile, type M. 

Figure D-38. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1970 to 12/31/1979, 

50th-percentile, type M. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-39. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 12/31/1988, 

50th-percentile, type M. 

Figure D-40. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1989 to 12/31/1993, 50thpercentile, type M.
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-41. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1994 to 12/31/2005, 50thpercentile, type M.

Figure D-42. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), from all intakes 1/1/1952 to 

12/31/2005, 50th-percentile, type M. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-43. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/1961, 

84th-percentile, type M. 

Figure D-44. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1962 to 12/31/1969, 

84th-percentile, type M. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-45. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1970 to 12/31/1979, 

84th-percentile, type M. 

Figure D-46. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 12/31/1988, 

84th-percentile, type M. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-47. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 01/01/1989 to 12/31/1993, 

84th-percentile, type M.

Figure D-48. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium

intake rates compared with measured plutonium-in-urine results (dots), 01/01/1994 to 12/31/2005, 

84th-percentile, type M. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-49. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), from all intakes 1/1/1952 to 

12/31/2005, 84th-percentile, type M. 

Figure D-50. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/1961, 

50th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-51. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1962 to 12/31/1971, 

50th-percentile, type S. 

Figure D-52. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1972 to 12/31/1979, 

50th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-53. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 12/31/1993, 

50th-percentile, type S. 

Figure D-54. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1994 to 12/31/2005, 

50th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-55. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/2005, 

50th-percentile, type S. 

Figure D-56. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/1961, 

84th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-57. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1962 to 12/31/1971, 

84th-percentile, type S. 

Figure D-58. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1972 to 12/31/1979, 

84th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-59. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 12/31/1993, 

84th-percentile, type S. 

Figure D-60. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1994 to 12/31/2005, 

84th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-61. Predicted plutonium bioassay results (line) calculated using IMBA-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/2005, 

84th-percentile, type S. 

Figure D-62. Predicted americium bioassay results (line) calculated using IMBA-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1972 to 12/31/1976, 

50th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-63. Predicted americium bioassay results (line) calculated using IMBA-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1977 to 12/31/1982, 

50th-percentile, type S. 

Figure D-64. Predicted americium bioassay results (line) calculated using IMBA-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1983 to 12/31/1988, 

50th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-65. Predicted americium bioassay results (line) calculated using IMBA-derived americium 

intake rates compared with measured americium lung burden results (dots) from all intakes 1/1/1972 

to 12/31/1988, 50th-percentile, type S. 

Figure D-66. Predicted americium bioassay results (line) calculated using IMBA-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1972 to 12/31/1976, 

84th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-67. Predicted americium bioassay results (line) calculated using IMBA-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1977 to 12/31/1982, 

84th-percentile, type S. 

Figure D-68. Predicted americium bioassay results (line) calculated using IMBA-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1983 to 12/31/1988, 

84th-percentile, type S. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-69. Predicted americium bioassay results (line) calculated using IMBA-derived americium 

intake rates compared with measured americium lung burden results (dots) from all intakes 1/1/1972

to 12/31/1988, 84th-percentile, type S. 

Figure D-70. Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/1961, 

50th-percentile, type SS. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-71. Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1962 to 12/31/1971, 

50th-percentile, type SS. 

Figure D-72. Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1972 to 12/31/1979, 

50th-percentile, type SS. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-73. Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 12/31/1993, 

50th-percentile, type SS. 

Figure D-74. Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1994 to 12/31/2005, 

50th-percentile, type SS. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-75. Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/2005, 

50th-percentile, type SS. 

Figure D-76. Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/1961, 

84th-percentile, type SS. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-77. Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1962 to 12/31/1971, 

84th-percentile, type SS. 

Figure D-78. Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1972 to 12/31/1979, 

84th-percentile, type SS. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-79. Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1980 to 12/31/1993, 

84th-percentile, type SS. 

Figure D-80. Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1994 to 12/31/2005, 

84th-percentile, type SS. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-81. Predicted plutonium bioassay results (line) calculated using IDOT-derived plutonium 

intake rates compared with measured plutonium-in-urine results (dots), 1/1/1952 to 12/31/2005, 

84th-percentile, type SS. 

Figure D-82. Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1972 to 12/31/1976, 

50th-percentile, type SS. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-83. Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1977 to 12/31/1982, 

50th-percentile, type SS. 

Figure D-84. Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1983 to 12/31/1988, 

50th-percentile, type SS. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-85. Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots) from all intakes 1/1/1972 

to 12/31/1988, 50th-percentile, type SS. 

Figure D-86. Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1972 to 12/31/1976, 

84th-percentile, type SS. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-87. Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1977 to 12/31/1982, 

84th-percentile, type SS. 

Figure D-88. Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots), 1/1/1983 to 12/31/1988, 

84th-percentile, type SS. 
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ATTACHMENT D

INTERNAL CO-EXPOSURE DOSIMETRY DATA (continued) 

Figure D-89. Predicted americium bioassay results (line) calculated using IDOT-derived americium 

intake rates compared with measured americium lung burden results (dots) from all intakes 1/1/1972 

to 12/31/1988, 84th-percentile, type SS. 
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ATTACHMENT E 

EVALUATION OF THE POTENTIAL FOR INTERNAL DOSE FROM NEPTUNIUM-237

TABLE OF CONTENTS

SECTION TITLE PAGE

E.1 Introduction .......................................................................................................................... 205

E.2 Operations ........................................................................................................................... 205

E.2.1 1962 to 1983 ............................................................................................................ 205

E.2.2 After 1983................................................................................................................. 206

E.3 Inventory Totals and Measured Quantities ........................................................................... 206

E.4 Measurements of Neptunium Waste at the Idaho National Engineering and 

Environmental Laboratory .................................................................................................... 207

E.5 Internal Dose Monitoring ...................................................................................................... 208

E.6 Workplace Air Monitoring ..................................................................................................... 208

E.7 Containment Measures During Operations .......................................................................... 208

E.8 Workplace Incidents............................................................................................................. 209

E.9 Shipment and Receipt Data ................................................................................................. 209

E.10 Potential for Exposures after 1983 ....................................................................................... 209

E.11 Conclusion ........................................................................................................................... 210

LIST OF TABLES

TABLE TITLE PAGE

E-1 Neptunium-237 fiscal-year-end inventories .......................................................................... 207

E-2 Mean masses and ratios of 239Pu and 237Np in waste drums ................................................ 208

E-3 Mean mass concentrations and ratios of 239Pu and 237Np in RFP waste drums .................... 208

E-4 Neptunium receipts from other sites..................................................................................... 209

E-5 Neptunium shipments from RFP to receiving sites ............................................................... 210
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EVALUATION OF THE POTENTIAL FOR INTERNAL DOSE FROM NEPTUNIUM-237 (continued)

E.1 INTRODUCTION

The following class of workers was added to the SEC (Sebelius 2013, p. 3):

All employees of the Department of Energy, its predecessor agencies, and their 

contractors and subcontractors who worked at the Rocky Flats Plant in Golden, 

Colorado, from April 1, 1952, through December 31, 1983, for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment or in 

combination with work days within the parameters established for one or more other 

classes of employees included in the Special Exposure Cohort.

This recommendation derived, in part, from the conclusion that neptunium processing at RFP took 

place at least until 1983, and that available monitoring data are insufficient for estimating potential 

internal exposures to neptunium. 

E.2 OPERATIONS

Neptunium Processing at the Rocky Flats Plant (Connor and Basso 1981, p. 11), states that 

processing included preparation of pure neptunium oxide, metal, and metal alloys as well as 237Np 

recovery from a variety of residues. Processes included dissolution, anion exchange, precipitation, 

filtration, calcination, conversion to fluoride, and reduction to metal. Fabrication steps such as casting 

and rolling were also sometimes performed for the production of high-purity metal shapes and foils. 

Neptunium was recovered from residual materials including sand, slag, crucibles, casting skulls, and 

various alloys containing plutonium, tin, uranium, or zirconium. This description of RFP neptunium 

operations is echoed in Actinide Processing at Rocky Flats (Connor and Basso 1984, p. 327).

E.2.1 1962 to 1983

Neptunium was processed at RFP as early as 1962 (Connor and Basso 1981, p. 11). There is no 

evidence of continuous routine neptunium operations. Rather, it points to a series of discrete tasks 

performed from 1962 through 1983 involving a few to a few hundred grams, usually processed by 

request from other DOE laboratories.

The first special-order request for neptunium processing at RFP came from Lawrence Radiation 

Laboratory, now known as LLNL, which required high-purity neptunium without gamma-emitting 

impurities. Techniques for purifying neptunium (including the removal of other actinides) were 

developed as a result of this project. The project also called for preparation of NpPu alloys, which 

were prepared by coreducing neptunium and PuF4 mixtures with calcium metal. This effort led to an 

ability to produce pure neptunium metal that could be cast with plutonium or uranium to form alloys. 

The first NpPu alloy was produced in 1964, and several kilograms of neptunium metal were produced 

in subsequent years for preparation of uranium and plutonium alloys.

Other specific projects involving 237Np processing included the preparation of high-purity neptunium 

oxide for the Oak Ridge National Laboratory (ORNL) isotope pool, neptunium metal foils for the 

Savannah River Site (SRS), and neptunium metal disks for use in the neutron dosimeter program for 

the liquid-metal, fast-breeder reactor (Connor and Basso 1981, p. 12).

Based on reviews of numerous reports in the SRDB as well as reviews of classified documents, it is 

evident that only relatively small quantities of 237Np were processed at RFP at any given time. The 
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EVALUATION OF THE POTENTIAL FOR INTERNAL DOSE FROM NEPTUNIUM-237 (continued)

largest single operation appears to have been a special order in 1966 that involved just over 300 g of 

neptunium (Byrne 1967).

E.2.2 After 1983

A single operation was described in Production-Scale Plutonium-Neptunium Separation and Residue 

Recovery at Rocky Flats Plant (Martella, Guyer and Leak 1987). This document does not make clear 

exactly when this operation took place, although it does indicate that the campaign lasted 

approximately 1 year. An interview of one of the authors, who was also the Principal Engineer who 

designed the process and directed the activities, estimated that the operation began around January 

1985 (ORAUT 2014a). 

This 1985 operation involved the processing of plutonium scrap containing as little as 0.5% (by 

weight) neptunium to separate and recover the two metals. The feed material for this process 

consisted of 63,728 g of plutonium containing 232 g of neptunium. The separation process involved 

oxidizing the plutonium residue, passing Pu(III) through an anion exchange resin bed, and leaving 

Np(IV) behind for subsequent elution, evaporation, denitrification, and calcination to oxide (Martella, 

Guyer, and Leak 1987, p. 10). The authors reported completion of 24 separations over the course of 

a year using this process, resulting in the purification of 58,282 g of plutonium by the removal of 222 g 

of neptunium (~96% recovery). NIOSH reinterviewed the coauthor (also Principal Engineer) who 

stated that project personnel consisted of five experimental operators who performed the glovebox 

work and the Principal Engineer (ORAUT 2014b).

The final purified plutonium that resulted from this operation contained only 0.0069% neptunium. The 

“neptunium product” consisted of 1,429 g of plutonium and 222 g of neptunium with a Pu:Np mass 

ratio of 6.4 (Martella, Guyer, and Leak 1987, p. 16). This operation involved no purified neptunium; 

the dose from any internal exposure would have been dominated by the overwhelming amount of 

plutonium in the mixture, making neptunium bioassay unnecessary. Given the much greater specific 

activity of 239Pu, plutonium, bioassay would account for all organ doses, of which plutonium would be 

the dominant component. An evaluation of the relative dose contribution from plutonium and 

neptunium from an inhalation intake of the PuNp mixture is presented later in Section E.10. 

According to the Principal Engineer, separation of plutonium and neptunium from alloys ended in 

1987 (ORAUT 2014a, p. 10). This statement is consistent with another reference that states that the 

neptunium program at RFP began in 1964 and was terminated in 1988 (Vejvoda 2005).

E.3 INVENTORY TOTALS AND MEASURED QUANTITIES

Documents and interviews with former employees indicate that neptunium was present at RFP from 

1962 to 2003. From 1963 through fiscal year (FY) 2003, neptunium quantities between 29 g and 

1,318 g are recorded in Material Balance Area accounts (Vejvoda 2005; Meyers 2013; DOE 2014a). 

Neptunium-237 inventory amounts at RFP from 1963 to 2003 obtained from the National Nuclear 

Security Administration Nuclear Materials Management and Safeguards System (NMMSS) are shown 

in Table E-1. Although neptunium work began in 1962, no inventory information has been found for 

the end of 1962 (the first year of production). The neptunium program was reportedly terminated in 

1988 (Vejvoda 2005), but neptunium remained in the RFP nuclear material inventory until 2003. 

Nuclear material holdup surveys were conducted sometime after 1990 in response to concerns by the 

Defense Nuclear Facilities Safety Board about unaccounted for nuclear material at Rocky Flats. 

These surveys identified gram quantities of residual neptunium in previously cleaned gloveboxes in 

Building 771, Rooms 153 and 182 (ORAUT 2014c, p. 11). 
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EVALUATION OF THE POTENTIAL FOR INTERNAL DOSE FROM NEPTUNIUM-237 (continued)

Table E-1. Neptunium-237 fiscal-year-end inventories.

a

FY

Np-237 

(g)

1963 29

1964 601

1965 1,292

1966 740

1967 1,215

1968 972

1969 1,190

1970 1,105

1971 1,318

1972 788

1973 768

1974 470

1975 485

1976b 485

FY

Np-237 

(g)

1976Ab 468

1977 458

1978 567

1979 492

1980 744

1981 486

1982 699

1983 869

1984 1,040

1985 931

1986 985

1987 995

1988 970

1989 935

FY

Np-237 

(g)

1990 937

1991 941

1992 941

1993 941

1994 941

1995 941

1996 935

1997 926

1998 930

1999 880

2000 900

2001 886

2002 636

2003c 501

a. Sources: Vejvoda (2005), Meyers (2013), DOE (2014a).

b. Inventory values reported for FY 1976 and FY 1976A correspond with a change in the

federal FY to begin October 1. The U.S. budget year began July 1 before this time

(ORAUT 2013a, p. 2; OMB 2012, p. 8). The values here agree with those reported

elsewhere (Meyers 2013) for the June 30 and September 30, 1976 quarterly inventory

periods.

c. Last reported inventory was on June 30, 2003.

It is difficult to draw firm conclusions about neptunium processing from the inventory table above. The 

data are presented primarily to show that the total neptunium inventory at RFP was always relatively 

low, typically less than a kilogram (compared to much greater inventories of plutonium for the same 

period). For example, the RFP site description TBD states that Building 771 recovery operations were 

processing 240 kg of plutonium daily by 1965. This daily plutonium throughput exceeds any single 

neptunium processing run by 2 to 3 orders of magnitude (ChemRisk 1992). A look at the year-to-year 

values reveals that inventory variability was higher in the earlier years and more stable after the mid1980s.

E.4 MEASUREMENTS OF NEPTUNIUM WASTE AT THE IDAHO NATIONAL ENGINEERING

AND ENVIRONMENTAL LABORATORY

As shown later in Table E-5, RFP shipped neptunium to several DOE sites, including neptuniumcontaining waste to the Idaho National Engineering and Environmental Laboratory, now known as the 

Idaho National Laboratory (INL). A recent interviewee (ORAUT 2014d) identified two volumes of a 

2004 document titled Validation of the Rocky Flats Plant Radionuclide Inventory in the Historic Data 

Task Using SWEPP Assay Data (Blackwood 2004; Blackwood and Hoffman 2004). Volume 1 of that 

document provides nondestructive assay (neutron and gamma) measurements of RFP waste stored 

at the INL Radioactive Waste Management Complex (RWMC). Table E-2 presents measurements 

showing that, in drums containing 237Np, 239Pu was also present at 239Pu:237Np mass ratios ranging 

from 105 to 6,450. Table E-3 presents measurements showing that the 239Pu:237Np mass 

concentration ratios ranged from 109 to 5,820. The predominance of 239Pu in the waste reflects the 

prevailing ratios in the RFP operating environment where neptunium was processed. Potential organ 

doses to workers packaging and handling this waste would, as previously discussed, be dominated by 

the plutonium contribution.
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EVALUATION OF THE POTENTIAL FOR INTERNAL DOSE FROM NEPTUNIUM-237 (continued)

Table E-2. Mean masses and ratios of 239Pu and 237Np in waste drums.a,b

Waste matrix 

Pu-239 

n

Pu-239 

mean (g) 

Pu-239 

σ (g) 

Np-237 

n

Np-237 

mean (g) 

Np-237 

σ (g) 

Pu-239/Np-237 

isotopic mass ratio 

Graphite 1,307 2.60 × 101 2.75 × 101 4 4.03 × 10-3 5.78 × 10-5 6.45 × 103

Filters 68 2.76 × 101 3.16 × 101 54 5.07 × 10-3 3.19 × 10-3 5.44 × 103

Mixed metals 523 1.46 × 101 3.14 × 101 14 1.82 × 10-2 3.64 × 10-2 8.02 × 102

First- and secondstage sludge3,095 6.17 8.74 281 5.86 × 10-2 8.01 × 10-2 1.05 × 102

a. Source: Graphite: Blackwood and Hoffman (2004, p. 75); filters: (p. 81); mixed metals: (p. 87); first- and second-stage

sludge: (p. 93).

b. Data in this table are compiled from measurements of radioactive waste stored at the INL RWMC. The symbols n and σ

are the number of measured drums and standard deviation, respectively, for the log-transformed data.

Table E-3. Mean mass concentrations and ratios of 239Pu and 237Np in RFP waste drums.a,b

Waste matrix

Pu-239

n

Pu-239

mean

(g/kg 

waste)

Pu-239

σ (g/kg 

waste)

Np237

n

Np-237

mean

(g/kg 

waste)

Np-237

σ (g/kg waste)

Pu-239:Np-237

isotopic mass 

concentration ratio

Graphite 1,307 3.67 × 10-1 4.16 × 10-1 4 6.31 × 10-5 5.78 × 10-5 5.82 × 103

Filters 68 1.06 1.03 54 2.96 × 10-4 3.80 × 10-4 3.58 × 103

Mixed metals 523 2.43 × 10-1 4.75 × 10-1 14 6.92 × 10-4 1.99 × 10-3 3.51 × 102

First- and secondstage sludge3,095 3.44 × 10-2 5.10 × 10-2 281 3.15 × 10-4 4.05 × 10-4 1.09 × 102

a. Source: Graphite: Blackwood and Hoffman (2004, p. 76); filters: (p. 82); mixed metals: (p. 88); first- and second-stage

sludge: (p. 94).

b. Data in this table are compiled from measurements of radioactive waste stored at the INL RWMC. The symbols n and σ

are the number of measured drums and standard deviation, respectively, for the log-transformed data.

E.5 INTERNAL DOSE MONITORING

Captured documents contain only two reported neptunium bioassays. A urinalysis result in June 1966 

was “Below Significant Level” (Hammond 1966a, p. 79; Dow 1966–1969, p. 8). A urinalysis result in 

July 1966 reported 0.9 dpm/24 hr; a followup body count showed no detectable uptake (Hammond 

1966b, p. 83; Dow 1966–1969, p. 10).

E.6 WORKPLACE AIR MONITORING

No neptunium-specific workplace air monitoring occurred.

E.7 CONTAINMENT MEASURES DURING OPERATIONS

The neptunium processing descriptions described in, Neptunium Processing at the Rocky Flats Plant

(Connor and Basso 1981) support the understanding that, by 1983, all neptunium operations were 

being performed in glovebox facilities. For example, the document states (Connor and Basso 1981, 

p. 12):

A typical glove box for aqueous Np processing consisted of a “wet” section (for aqueous 

processes) and a “dry” section (for calcining precipitates and weighing powders) 

separated by an air lock. Each section had separate air inlet and exhaust filters. A door 

in the air lock was used to pass equipment and material between the wet and dry 

sections. A 1⁄8-inch-thick lead sheet was bonded to the stainless steel portion of the glove 

box and 1⁄4-inch leaded glass was placed over the glove box windows as a shield against 

gamma radiation from the neptunium and its impurities.
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ATTACHMENT E

EVALUATION OF THE POTENTIAL FOR INTERNAL DOSE FROM NEPTUNIUM-237 (continued)

Only one neptunium processing operation is known to have occurred after 1983. It involved the 

separation of neptunium from plutonium residues in 1985 (see Section E.3). According to the 

Principal Engineer who designed the process and directed activities, this operation was performed in 

gloveboxes and tanks. The tanks contained feed material (plutonium and neptunium nitrate solution) 

that was piped directly into the gloveboxes (ORAUT 2014b).

E.8 WORKPLACE INCIDENTS

No radiological incidents at RFP involving neptunium have been identified.

E.9 SHIPMENT AND RECEIPT DATA

NMMSS records of neptunium shipments and receipts by RFP from 1983 until site closure in 2003 

(DOE 2014b to 2014e) were obtained and analyzed. Table E-4 shows the FYs during which gram 

amounts of neptunium were received at RFP from Los Alamos National Laboratory (LANL), SRS, 

ORNL, and LLNL. No material was received after FY 1986. The three-letter designations in 

parentheses after the site acronyms are the DOE Reporting Identification Symbol codes for the sites. 

Neptunium shipments from RFP to other DOE sites are shown in Table E-5.

Table E-4. Neptunium receipts from other sites (g). 

FY

LANL 

(AUA)

SRS 

(DZA)

ORNL 

(FZC/FZG)

LLNL 

(LZB)

1983 0 556 199 105

1984 0 0 199 1

1985 99 0 0 228

1986 1 113 0 0

E.10 POTENTIAL FOR EXPOSURES AFTER 1983

From 1962 through 1983 there existed the potential for unmonitored worker exposures to 237Np. 

There was insufficient data available to estimate intakes of 237Np during that period. 

After 1983, only one operation involving neptunium (Section E.3). Workplace monitoring for this 1985 

project included continuous air monitors (CAMs) and contamination surveys typical of the plutonium 

processing areas. Radiation Monitors (technicians) were available to provide radiological safety 

support. Personnel involved in the project were on bioassay programs typical of the plutonium 

processing environment (i.e., routine urinalyses and body counts) (ORAUT 2014b).

The previous discussion noted that this operation did not involve purified neptunium, but rather 

plutonium with neptunium as a contaminant. Even the most highly concentrated neptunium product 

produced by this separation process was still mostly plutonium, with a Pu:Np mass ratio of 6.4. 

Because the specific activity of 239Pu is about 90 times greater than that of 237Np, the activity ratio of 

this PuNp mixture is greater than 500. As a result, all organ doses from an intake of the mixture 

would be dominated by the plutonium component. The neptunium dose component would contribute 

only about 0.1% for type M plutonium for most organs, and only about 1% for type S plutonium 

(ORAUT 2014e).
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EVALUATION OF THE POTENTIAL FOR INTERNAL DOSE FROM NEPTUNIUM-237 (continued)

Table E-5. Neptunium shipments from RFP to receiving sites (g).a

Fiscal 

year

LANL 

(AUA)

SRS 

(DZA)

ORNL 

(FZC/FZG)

Hanford 

(HRA)

LLNL 

(LZB)

Nevada Test Site 

(NAB)

WIPP 

(VPA)

INL Waste 

(VSB)

1983 0 0 378 0 220 40 0 0

1984 0 0 0 0 159 0 0 0

1985 0 0 0 0 146 26 0 22

1986 73 0 0 2 0 0 0 0

1987 1 0 0 0 0 0 0 0

1988 2 8 0 0 0 0 0 0

1989 0 36 0 0 0 0 0 0

1990 0 0 0 0 0 0 0 0

1991 0 0 0 0 0 0 0 0

1992 0 0 0 0 0 0 0 0

1993 0 0 0 0 0 0 0 0

1994 0 0 0 0 0 0 0 0

1995 0 0 0 0 0 0 0 0

1996 0 0 0 0 0 0 0 0

1997 5 0 0 0 7 0 0 0

1998 0 0 0 0 0 0 0 0

1999 0 46 0 0 0 0 1 0

2000 0 18 0 0 0 0 0 0

2001 0 0 0 0 0 0 174 0

2002 0 297 0 0 0 0 324 0

2003 0 394 0 0 0 0 11 0

a. WIPP = Waste Isolation Pilot Plant.

As mentioned previously, interviews with the Principal Engineer and coauthor of the 1987 document 

Production-Scale Plutonium-Neptunium Separation and Residue Recovery at Rocky Flats Plant

(Martella, Guyer, and Leak 1987) indicated that everyone who worked in a “hot” building had a 

urinalysis every few months as well as annual whole-body counts (ORAUT 2014a, 2014b). If this 

individual or any of his coworkers had received neptunium intakes from this operation, the resulting 

organ doses would be adequately accounted for by the plutonium bioassay program. Plutonium 

bioassay data are readily available for RFP claimants.

No other specific operations involving neptunium were identified, other than the 1985 operation. If 

other operations of a similar nature (i.e., neptunium associated with plutonium) did take place, 

plutonium bioassay would have similarly accounted for any doses due to neptunium.

E.11 CONCLUSION

In conclusion, there is no evidence that 237Np intakes occurred at RFP after December 31, 1983. If 

intakes had occurred during this period from the one known neptunium operation, the resulting organ 

doses would be adequately accounted for by the available plutonium bioassay data. 
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ATTACHMENT F 

REPORTED PRESENCE OF MAGNESIUM-THORIUM ALLOY

TABLE OF CONTENTS

SECTION TITLE PAGE

F.1 Introduction .......................................................................................................................... 212

F.2 Magnesium-Thorium Search Results ................................................................................... 212

F.3 Conclusion ........................................................................................................................... 215
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ATTACHMENT F

REPORTED PRESENCE OF MAGNESIUM-THORIUM ALLOY (continued)

F.1 INTRODUCTION

Workers at RFP have provided statements that MgTh alloy plates were used, handled, or processed 

on site at RFP. One worker stated (Kusmierczak 2007):

Earlier this month a former Rocky Flats worker related to me thru a third party (who wants 

to remain anonymous) information concerning the use of magnesium thorium alloy plates 

at Rocky Flats. You may remember that Dow workers submitted affidavits that Dow 

shipped these plates to Rocky Flats.

The information relayed to me was, "They were brought in on the 903 pad send [sic] to 

881 to refine then sent to the MOD center for modification to fit the Semi Trucks as to 

make them bullet proof. A Semi bed was brought in stripped down and the sheets were 

refined to help armor plate the trucks for transport of materials. He also said they were 

used in the Train cars for support!"

The affiant indicated that work was performed with this thorium material at Dow Chemical Company in 

Madison, Illinois, from 1962 to 1975 but did not provide any specific information on when the 

shipments might have been made to RFP. No corroborating evidence for the assertion that MgTh 

alloys were used or present at RFP was found during the research into MgTh alloy efforts in the DOE 

complex over the applicable period. RFP personnel who would have been aware of the receipt of 

these types of materials were interviewed and none were aware that MgTh alloy was ever present or 

used in any significant quantity at RFP (SC&A 2007a, pp. 66–67).

Two other Colorado facilities in the proximity of RFP handled nuclear materials and specifically 

handled MgTh alloys, nuclear weapons, or related equipment that would have used MgTh alloys in 

their designs (Rocky Mountain Arsenal and Lockheed Martin-PJKS Test Facility). The Dow Madison 

individual who had stated that MgTh alloys were sent from Madison to RFP was interviewed and 

asked about the possibility that the MgTh could have been shipped to one of the other locations rather 

than RFP. The individual indicated that this was a possibility and that he was not aware of the 

existence of the other two facilities. 

F.2 MAGNESIUM-THORIUM SEARCH RESULTS

An initial search of documents was conducted to determine if a link could be established between 

MgTh alloy and RFP. This search included all listed sites and identified the following documents 

containing information on MgTh alloy: 

• Owen 2007. A Freedom of Information Act (FOIA) request “for records that pertain to 

shipments of thorium-magnesium alloy metal between the Dow Chemical Company at its 

Madison, Illinois, facility or at the Walnut Creek-Pittsburgh, California, dual facility and the 

Rocky Flats nuclear weapons facility in Colorado between 1951 and 1973 [emphasis added]” 

resulted in a response from the DOE Office of Legacy Management identifying 10 shipping 

orders (Dow 1957a to 1957d, 1958a to 1958d, 1959a, 1959b). The identified documents, 

however, are all for shipments of plutonium samples from RFP (Dow) to SRS (DuPont) from 

1957 to 1959. 

• DOE 2007. SRS responses to FOIA requests that contain request from an East Alton, Illinois 

attorney about MgTh alloy metal.
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ATTACHMENT F

REPORTED PRESENCE OF MAGNESIUM-THORIUM ALLOY (continued)

• SC&A 2007b. The SC&A review of the evaluation report for SEC-00030 (NIOSH 2006) for 

RFP has no estimate of dates when MgTh might be present at Rocky. Their finding (p. 18) 

states, “It is clear from NUREG-1717 and the other considerations presented above that 

knowledge of the approximate quantities, periods, and processing status of the MgTh alloy is 

needed before any reliable conclusions can be arrived at about doses to RFP workers from 

this material [emphasis added].”

• Author unknown undated a. Production of Magnesium Thorium Alloy, 1954-73, with 

production data from the Minerals Yearbook Metals and Materials, indicates that Dow Madison 

produced MgTh alloy from 1954 to 1965 and from 1967 to 1973 [emphasis added].

• SC&A 2007b. A Focused Review of Operations and Thorium Exposures at the Dow Chemical 

Company Madison Plant, in which four former RFP managers state that MgTh alloy was never 

used in any process for development or fabrication of weapons components.

• SRGA 2008. The minutes of the Advisory Board on Radiation and Worker Health (ABRWH) 

Meeting 56 in June 2008 on the Dow Madison SEC petition state that DOE concluded that 

Dow made MgTh alloys HK31A and HM21A for use in nuclear weapons in 1957 and 1958 (p. 

32) and that MgTh alloys were used in atomic weapons from 1956 to 1969 (p. 38) [emphasis 

added].

Additional research and reviews were conducted on topics potentially related to the presence of MgTh 

alloys at RFP. Searches were performed with key words that included the terms “magnesium” and 

“thorium” as well as their acronyms. These searches identified further search terms, including HK-31 

and HK-31A (codes used to identify MgTh). These additional searches identified the following 

documents about MgTh operations at Dow Madison in Illinois, Wyman-Gordon in Massachusetts, and 

the Kansas City Plant in Missouri, but none of these relates to RFP:

• Silverstein 1957 (Dow Madison). Dow memorandum about firefighting concerns associated 

with MgTh alloys.

• Peloubet 1959 (Dow Madison). Letter from Dow to the U.S. Atomic Energy Commission 

(AEC) about ventilation for MgTh welding.

• Silverstein 1958 (Dow Madison). Dow spectrographic analysis results for dust samples 

obtained from sanding HK-31.

• Mitchell 1958b (Dow Madison). Analytical results for air samples obtained during hand 

sanding HK-31.

• Levy 1957 (Dow Madison). Dow memorandum making a correction to a previous letter about

exposure to HK-31.

• Dow 1957e (Dow Madison). Dow memorandum explaining the new federal law passed in 

January 1957 that placed a limit on the external radiation a worker could receive.

• Mitchell 1957 (Dow Madison). Dow memorandum documenting sample turnover of two dust 

samples collected during hand sanding of HK-31.
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REPORTED PRESENCE OF MAGNESIUM-THORIUM ALLOY (continued)

• Mitchell 1958 (Dow Madison). Dow memorandum documenting sample turnover of three dust 

samples collected during sanding of HK-31 with vertical metallurgical sanding discs.

• Author unknown undated b (Dow Madison). Undated description of the composition, 

hardness, and strength of HK-31A material.

• GZA 1994 (Wyman-Gordon). Report on a 1993 survey of MgTh residual contamination on the 

Wyman-Gordon site.

• Various 1961–1975 (Kansas City Plant). A variety of operational documents from the 1960s 

and 1970s about MgTh alloy and a Bendix association to Dow Madison. 

• BAC 1970 (Kansas City Plant). Two memoranda discussing fabrication of MgTh units to be 

machined as part of the Radiation Detection Capability Program. 

• BAC 1972 (Kansas City Plant). Two memoranda listing personnel who had received 

physicals, orientation, or clearances to work with MgTh waste.

• BAC 1957–1970 (Kansas City Plant). Document specifying the health and safety procedures 

to be followed when handling MgTh (2% Th).

An additional site visit and data capture was performed at the Environmental Management 

Consolidated Business Center (EMCBC) in Denver, Colorado. It included classified document 

reviews and interviews and generated the following MgTh results that do relate to RFP:

• ORAUT 2013b. EMCBC documents on the safe, secure transport (SST) and safe, secure 

railcar from the Atomic Materials Rail Transport Car system Modification Center, Books 1-4 

(documents classified); classified document reviews for MgTh with the unclassified and 

released notes indicating no related information was identified.

• ORAUT 2013c. EMCBC interview; discussed SST and SSR work and provided photos of 

related operations; performed as a classified interview with the unclassified and released 

notes indicating that no MgTh-related information was identified.

• ORAUT 2013d. EMCBC railcar procedures; classified document reviews for MgTh with the 

unclassified and released notes indicating no related information was identified.

• ORAUT 2013e. EMCBC documents on the SSR; classified interviews for MgTh with the 

unclassified and released notes indicating no related information was identified.

• ORAUT 2013f, 2013g; NIOSH 2013b. Interviews about MgTh and any related operations. No 

related information was identified.

• NIOSH ca. 2013. Interview about MgTh operations. No related information was identified.

• Author unknown 1976b. RFP thorium shipment and receipt records for 1952 to 1976 with no 

indication or support of MgTh operations.

As indicated in the interviews, the armor or shielding plate installation operation associated with the 

potential use of MgTh alloys was performed at a specific location at RFP. Individuals who were 
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ATTACHMENT F

REPORTED PRESENCE OF MAGNESIUM-THORIUM ALLOY (continued)

directly involved in the work were interviewed about the use of MgTh alloys. In addition, 

documentation and procedures associated with the operation were reviewed. While most of the 

information and documentation remains classified, the information that was released based on the 

reviews does not corroborate the use of MgTh alloys at RFP for this or any other operations.

Followup interviews and information indicates that RFP might have collaborated with Sandia National

Laboratories (SNL) on the design of the armor plating installed on vehicles and equipment at RFP. 

Because SNL documentation about the armor plating could contain information on plating 

composition, a keyword search of classified and unclassified SNL documents for anything about this 

operation was performed. SNL-Albuquerque and SNL-Livermore indicated that no related 

documentation was discovered in their searches. 

F.3 CONCLUSION

Based on the research for this assessment, the ORAU Team concludes there is no corroborating 

evidence for the use of MgTh at the RFP site. It is likely that there is confusion between RFP and 

other Denver-area sites, as well as confusion about MgTh alloy plates and other similar materials at 

RFP. All of the available information for MgTh alloys for other sites falls within the period of the SEC 

class for RFP (1952 to 1983). Therefore, if any undocumented MgTh operations did occur at RFP 

that related to these other sites’ operations, the RFP operational dates would be covered under the 

existing RFP SEC class.
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ATTACHMENT G 

TRITIUM ISSUES

TABLE OF CONTENTS

SECTION TITLE PAGE

G.1 Introduction .......................................................................................................................... 217

G.2 Monitoring Data.................................................................................................................... 217

G.3 Dose Assignment for 1973................................................................................................... 221

G.4 Dose Assignment After 1973................................................................................................ 221

LIST OF TABLES

TABLE TITLE PAGE

G-1 Reported tritium air concentrations from the August 30, 1974, release................................. 219
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ATTACHMENT G

TRITIUM ISSUES (continued)

G.1 INTRODUCTION

The potential for tritium exposure to RFP personnel was not considered significant until an 

unexpected release occurred in April 1973. Because tritium monitoring was not rigorous before this 

event, a tritium bounding method was developed for this analysis that uses information from the 1973 

tritium incident as the maximum exposure scenario. 

G.2 MONITORING DATA

Although tritium was used as a boost gas in weapons and as target material in neutron generators, it 

was not processed or handled in significant quantities at RFP. Tritium was monitored in the 

environment around the site for a time, but that monitoring ceased and was left to the State of 

Colorado for a brief period before an environmental release that occurred in April 1973. No analytical 

records have been found that might help establish the RFP workplace tritium environment before 

1973.

The management of RFP responded quickly to the potential for tritium workplace and environmental 

contamination with the release in April 1973 of 500 Ci to 2,000 Ci of tritium, primarily from 

Building 779A, and its eventual detection in waters draining into a reservoir serving as a municipal 

drinking water supply (AEC 1973; Arkell 1976). The release did result in tritium exposure to a small 

number of RFP personnel. Subsequent workplace monitoring and personnel bioassay was 

implemented, in part to establish the baseline tritium environment against which future incidents could 

be evaluated. A smaller and less impactful tritium release occurred in September 1974 from 

Building 777. The subsequent investigation report (AEC ca. 1974) includes release details along with 

summaries of tritium workplace monitoring results before the incident for comparison. These data 

provide the basis for a model for bounding chronic tritium exposures to workers and of smaller, 

less-notable tritium releases that might have occurred before 1973.

Several factors single out the 1973 tritium release as bounding for the entire history of RFP 

operations. These factors include the large quantity of tritium involved, the chemical form of the 

released tritium, and the meteorological conditions at the time of the release. Other documented 

releases involved smaller quantities of elemental tritium having a much smaller DCF than the tritium 

oxide released in 1973. Bounds for personnel tritium exposures after the 1973 release can be 

developed based on measurement results from the personnel bioassay, air sampling, and workplace 

contamination monitoring data for tritium, which became more common after that release. There are 

only very limited tritium measurement results before 1973 because tritium was not perceived as a 

radionuclide of occupational or environmental interest at RFP. Methods for bounding tritium 

exposures before 1973 are more difficult to develop as a result of this lack of measurement data.

According to a ChemRisk report (1994), there was no environmental monitoring for tritium before 

1970, and little in the way of workplace monitoring until after the 1973 tritium release; therefore, 

evidence of tritium releases before 1973 is primarily anecdotal. There was a 600-Ci release of 

elemental tritium from a different source in 1968. However, that release is not related to the tritiated 

plutonium shipments. The report said the following about possible releases from tritiated plutonium 

shipments (ChemRisk 1994, p. 285):

The 1973 findings associated with the tritiated plutonium initiated an investigation of 

other possible similar shipments and processing of tritiated plutonium. The investigation 

discovered three other shipments with maximum estimated tritium releases of 57 Ci 

(April 1969), 40 Ci (March 1971), and 29 Ci (November 1971).
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ATTACHMENT G

TRITIUM ISSUES (continued)

The reported investigation and the documented 1968 release of elemental tritium are the only sources 

of information about other possible releases. The 1968 release was elemental tritium with no 

significant environmental or personnel exposure. None of the three identified potential releases from 

tritiated plutonium was near the magnitude of the 1973 release. There is no evidence of a tritium 

release comparable to the magnitude and effects of the 1973 release before that year. 

Despite the lack of measurement data, it is possible to develop tritium exposure bounds for before 

1973 based on measurement results in a Rocky Flats Area Office (RFAO) report after a tritium 

release in one of the RFP production buildings on August 30, 1974 (AEC ca. 1974). The information 

in this report includes measurement data (i.e., results from air samples, surface contamination 

surveys, and bioassay) from the production area where the release occurred as well as comparison 

data from other areas before, during, and after the release. Several factors support the use of these 

data as surrogates for bounding the tritium environment at RFP before 1973:

• Background tritium levels immediately before the incident, although undoubtedly elevated 

since the more significant 1973 release, were well below dosimetrically significant values and 

can be considered as fairly representative of typical background levels for this analysis. The 

background tritium levels monitored in the months before the 1974 incident are consistent with 

internal radiation doses from tritium of well under 1 mrem annually. They are dosimetrically 

insignificant in this sense.

• The quantity of released tritium (1.5 Ci) was significantly less than that in 1973 and was 

probably more typical of potential undocumented releases in work areas, particularly those 

from opening contaminated shipping containers.

• The 1974 1.5-Ci tritium release is the only documented release from a shipping container in 

the RFP workplace. It is taken to be typical because there are no other such documented 

releases to use in forming the model. There is documented concern about tritium releases, as 

noted by ChemRisk (1994, p.38):

As early as 1962, Rocky Flats maintained instruments for detection of tritium gas 

in particular work areas of the plant because operations have sometimes resulted 

in the storage of tritium containers.

The instruments available to RFP at that time were only semi-quantitative for indicating the presence 

of tritium; NIOSH has found no records of these results. Because NIOSH has only identified six 

documented releases from 1968 to1974 (an average of 1 per year), the application of a daily release 

would be a significant, and therefore bounding, overestimate of the number of RFP tritium releases.

• Tritium was released to the workplace environment rather than in a glovebox.

• The release involved elemental tritium (HT) but not tritium oxide [as tritiated water vapor 

(HTO)].1

 

1 The impact of the 1973 tritium release was largely due both to the quantity (500 Ci to 2,000 Ci) and the chemical form 

(HTO) of the material. The presence of tritium oxide in the 1973 release resulted from peculiarities of the plutonium 

recovery operation from which it came. There is no indication that any other tritium release at RFP involved the oxide. 

Tritium in its elemental form (HT, T2) is far more likely to have been a contaminant because of the nature of its possible 

RFP source terms: tritiated accelerator targets (neutron generators), plutonium hydride in recovery operations, and 

boost gas in returned reservoirs or pits.
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ATTACHMENT G

TRITIUM ISSUES (continued)

• The tritium was released from a contaminated shipping container that was procured by RFP in 

1970 and can be taken as representative of shipping containers in use before 1973.

• The incident occurred close enough in time to the 1973 tritium release that work practices and 

controls were likely more similar to those before 1973 than to those even 1 or 2 years later, as 

procedures and controls evolved with greater sensitivity to the potential for tritium 

contamination.

The RFAO report provides the best source of monitoring data for use in bounding both chronic and 

accidental tritium exposures to RFP personnel before the unique circumstances of the 1973 release. 

The report states that elevated tritium concentrations were detected in air samples from Room 452 

(Special Assembly Area) in Building 777 and from the Building 205 exhaust plenum servicing Building 

776/777 from August 30 to September 4, 1974 (AEC ca. 1974, p.36). Subsequent sampling and 

investigation of the elevated sample results concluded that about 1.5 Ci of tritium was released from 

the exhaust system of Room 452, Building 777, when a shipping container (referred to as a “pressure 

cooker”) received in July 1974 was opened on a downdraft table in Room 452 on August 30 (AEC ca. 

1974, pp. 36–39). No elevated environmental tritium levels were detected as a result of the incident, 

but workplace tritium levels 7 times the applicable radioactivity concentration guide were detected in 

air samples collected on August 30 in Room 452 adjacent to the downdraft table, with average 

concentrations for the work week about 1.5 times the guide. Table G-1 below shows the reported 

values.

Table G-1. Reported tritium air concentrations from the August 30, 1974, 

release (μCi/m3

).a 

Sampling reference

Plenum 205,

Building 776/777b

Room 452,

Building 777

Normal concentrations <1 × 10-2 <1 × 10-2

August 29–30, 1974 0.148 37.7

September 3–4, 1974 2.51 1.1

a. Source: AEC (ca. 1974, pp. 9, 93–96). 

b. The Special Assembly Glovebox Line in Room 452, Building 777, was normally served by 

Plenum 206, but exhaust air from this area was vented through Plenum 205 from 

February 11 to August 7, 1974, while a new Plenum 206 was constructed. A tritium air 

sampler for Plenum 206 was installed on August 30, 1974, but showed no elevated 

results. However, both plenums showed elevated removable tritium contamination (AEC 

ca. 1974, pp.74-82).

Air Sample Results

Results from daily air samples in Room 452, Building 777, are available from June 3 to September 11, 

1974. The air sampler was near the downdraft table entry to the Special Assembly Line where the 

tritium-contaminated “pressure cooker” was opened, and it was the only tritium air sampler in 

Building 777 at the time. Room air samples were collected in a water bubbler during the day shift 

(approximately 6 to 8 hours sampling time) at an air flow rate of 2 L/min.

The average and standard deviation of daily air sample results before August 30, the day of the tritium 

release from the contaminated shipping container, are 5,343 ±4,518 pCi/m3 (AEC ca. 1974, pp. 87–

89). The result on August 30 is 37,676,609 pCi/m3

, and the sample taken on September 3 indicated a 

tritium concentration in the room air of 1,098,901 pCi/m3

. However, the September 3 result is suspect 

because the sample was collected in the same vessel that was used on August 30, which had not 

been cleaned. Smear surveys of Room 452 on September 3 failed to show significant tritium 

contamination (AEC ca. 1974, pp.37–38). Tritium levels in Building 777 were known to be somewhat 
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ATTACHMENT G

TRITIUM ISSUES (continued)

elevated over normal background because of residual contamination present since the 1973 tritium 

release.

Bioassay Results

The practice of pulling a sample of air from within shipping containers through a tritium air monitor to 

check for contamination was implemented after the 1973 tritium release. This practice was 

discontinued after urinary tritium results in the range of 0.75 to 1.3 μCi/L were detected in May 1974 

for the [redacted] who performed the monitoring. The [redacted] urinary tritium dropped to less than 

0.1 μCi/L beginning in early July 1974 (AEC ca. 1974, pp. 18–19).

Everyone who worked in Room 452, Building 777, submitted urine samples after the August 30 tritium 

release, with a high result of 32,320 pCi/L (AEC ca. 1974, p. 90).

The report indicates that both a Denver resident and a RFP worker who did not work in radioactive 

material-handling areas were sampled with results <0.01 μCi/L (<10,000 pCi/L) (AEC ca. 1974, p.30). 

The Denver resident is identified in Table A1-3 of that report, and Worker ID [redacted] is believed, by 

implication, to be the RFP nonradiological worker.

Work Area Smear Surveys

Over 200 smear results for tritium are tabulated in the RFAO report (AEC ca. 1974, pp. 74–82). Most 

appear to be surveys inside gloveboxes, but there are also workplace area results that can be used 

as indicators of likely sources of internal contamination of workers after an event such as that in 

August 1974. There were 42 workplace smear results. They range from <100 pCi/smear to 1 × 107

pCi/smear, with an average of about 8.2 × 104 pCi/smear.

The exhaust plenums and the Kathabar air driers (which use a lithium chloride solution called 

Kathene) appear to have collected the greatest amount of tritium after the release. Workers 

responsible for changing filters in the plenums or recharging the Kathabar systems would appear to 

be at greatest risk for tritium uptake after the initial release.

Assessment of the 1974 Incident

The 1.5-Ci tritium release from a contaminated shipping container occurred on August 30, 1974. The 

RFAO report provides air survey, bioassay, and smear survey results (AEC ca. 1974). Specific urine 

sample collection dates were not included in the report, but data were matched to two NOCTS claims 

that reported a collection date of September 5, 1974. A dose assessment was performed assuming 

an intake date of August 30, 1974, and the largest reported bioassay result collected after the 

incident, 36,320 pCi/L. There was a slight discrepancy (one digit) between the result in the RFAO 

report and that in the NOCTS case file. The NOCTS value is assumed to be correct because it is the 

handwritten urinalysis record and is also the larger of the two values. The resulting dose is <1 mrem 

(0.15 mrem). 

Conclusion

The RFAO report of a 1.5-Ci tritium release on August 30, 1974, from a contaminated shipping 

container (“pressure cooker”) provides air survey, bioassay, and smear survey results that can be 

used to model similar releases (AEC ca. 1974). Assuming an event like this one occurred every 

workday of every year (i.e. 250 times a year), the resulting annual dose would be 37.5 mrem/yr. This 

should be assigned as a bounding estimate for all workers’ potential tritium exposure before 1973.
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ATTACHMENT G

TRITIUM ISSUES (continued)

G.3 DOSE ASSIGNMENT FOR 1973

The Investigation of the Tritium Release Occurrence at the Rocky Flats Plant (AEC 1973) describes 

the 1973 incident, which prompted the site to sample a number of workers for tritium exposure. A 

shipment of scrap plutonium from LLNL was discovered to have been contaminated with tritium. This 

material was processed at RFP from April 9 to 25, 1973, in Building 779A. Because it was not 

immediately identified as being contaminated, monitoring of potentially exposed individuals did not 

begin until late September 1973.

Two-hundred-fifty people were sampled after the discovery; this included all personnel who worked in 

areas in which the contaminated scrap was processed or who were involved in the processing of 

wastes from this scrap. Due to the large sample load, raw urine samples were first analyzed in many 

of the cases. It was noted that the counting efficiency was only about 3% for these analyses, and that 

the corrections made for spectral shift could lead to abnormally high readings. Nineteen workers were 

initially identified as having elevated tritium levels in their urine. These samples were distilled and 

reanalyzed, which found that fourteen of these workers had levels below the 10,000 pCi/L action level 

established by the site. The [redacted] most exposed individuals were identified and details of their 

potential exposures, including bioassay results, are included in the above-cited investigation report. 

[redacted] of these [redacted] individuals can be found in NOCTS.

Tritium contamination was associated with plutonium scrap material; therefore, 3

H doses should be 

assigned to all individuals who were monitored for plutonium in 1973. Doses should be assessed on 

an individual basis using reported 3

H bioassay results and any additional information in the NOCTS 

file. For those who were not monitored for 3

H, dose should be assigned based on assumptions that 

are favorable to the individual claimant.

The tritium release incident report (AEC 1973) discusses the tritium sampling of 250 people as of 

October 15, 1973. There are no results or specific sample dates in the report for those individuals 

who did not exceed the 3

H bioassay action level of 10,000 pCi/L. Only [redacted] individuals were 

found to have results exceeding the action level, so this current assessment assumes that they are 

the maximally exposed workers in the incident and that those who were not monitored would not have 

been exposed at those same levels. After discussions about the tritium model at the RFP ABRWH 

Working Group meeting, NIOSH decided to adopt the use of the tritium model described in 

Richardson and Dunford (2001) for this assessment.

Based on this information, the following assumptions were applied:

• 3

H was in the form of HTO;

• Model for inorganic 3

H is described by Richardson and Dunford (2001);

• Mode of intake was injection (for modeling with IMBA);

• Intake date was April 9, 1973, the first day the material was processed;

• Sample was collected on October 15, 1973; and

• Result was 10,000 pCi/L (14,000 pCi/d).

Using this information, the total dose is 103 mrem. 

G.4 DOSE ASSIGNMENT AFTER 1973

The ORAU Team performed a co-exposure study using data from NOCTS for 1974 and 1975 in 

accordance with ORAUT-OTIB-0075, Use of Claimant Datasets for Coworker Modeling (ORAUT 
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ATTACHMENT G

TRITIUM ISSUES (continued)

2016). There are 38 individuals with tritium data in 1974 and 37 in 1975. When assessing tritium 

intakes for most sites, it is assumed that intake potential existed only for these individuals with tritium 

data in 1974 and 1975.

When assessing tritium intakes for most sites, it is assumed that intake potential existed only while 

tritium bioassay monitoring was being performed because monitoring is cheap, easy, and requires 

only spot samples, which is less of a burden than other forms of bioassay on both the employer and 

the worker. Because tritium was not of primary concern at RFP and was present only as a potential 

contaminant on equipment, a given individual was not placed on a routine sampling program. 

Instead, a program was established whereby one-tenth of the urine samples collected for plutonium 

analysis were also analyzed for tritium content (Bowman 1974) as well as the collection of samples 

when a particular concern was identified. Samples available in NOCTS for these 2 years indicate that 

analyses were performed throughout the year, with most individuals sampled only once.

For the purpose of the co-exposure study, it was assumed that each worker had the potential to be 

exposed at a constant level throughout the year in which the urine sample was collected. The 95th 

percentile was used because one-tenth of the population was sampled. The co-exposure study for 

1974 to 1975 yielded doses of much less than 1 mrem for everyone. 

For the years after 1975, there are fewer than 11 claims in NOCTS with tritium data, which is 

insufficient for performing a co-exposure study. Results for these years are consistent with those 

from the previous years and show a generally decreasing personnel exposure trend. 

Therefore, after 1973 no additional unmonitored dose due to tritium should be assigned to RFP 

workers.
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ATTACHMENT H 

CRITICAL MASS LABORATORY

TABLE OF CONTENTS

SECTION TITLE PAGE

H.1 Introduction .......................................................................................................................... 225

H.2 Personnel Exposure Concerns............................................................................................. 225

H.3 History of Uranyl Nitrate Criticality Experiments ................................................................... 225

H.4 Reevaluated Fission and Activation Product Level Estimates .............................................. 225

H.4.1 Thermal Power and Fission Rate Estimates in Bare Metal Experiments................... 226

H.4.2 Thermal Power and Fission Rate Estimates in Metal-Plus-Solution Experiments ..... 226

H.4.3 Thermal Power and Fission Rate Estimates in Solution-Only Experiments............... 226

H.4.4 Conclusions about Reactor Power and Fission Rates in Experiments ...................... 227

H.5 Surface Contamination......................................................................................................... 228

H.5.1 Radiological Survey Results ..................................................................................... 228

H.5.2 Maximum Reported Values ...................................................................................... 228

H.5.3 Evaluation of Removable Contamination Survey Results ......................................... 228

H.5.4 Discovery of Sealed Strontium-90 Sources............................................................... 233

H.5.5 Conclusions About Surface Contamination............................................................... 233

H.6 Workplace Air Monitoring ..................................................................................................... 235

H.6.1 Workplace Air Monitoring Requirements................................................................... 235

H.6.2 Air Sampling Locations............................................................................................. 236

H.6.3 Air Sampling Results ................................................................................................ 237

H.6.4 Conclusions About Workplace Air Monitoring ........................................................... 240

H.7 Assessment of Unmonitored Radiation Dose ....................................................................... 240

H.7.1 Calculation of Fission and Activation Product Content of Uranyl Nitrate Solution...... 241

H.7.2 Calculation of Inhalation Intakes and Committed Organ Doses ................................ 242

H.8 Summary and Conclusions .................................................................................................. 245

H.9 Data Sources ....................................................................................................................... 245

LIST OF TABLES

TABLE TITLE PAGE

H-1 Average reactor power and fission rates derived from measurements at CML ..................... 228

H-2 Maximum removable alpha contamination ........................................................................... 229

H-3 Maximum dose rates in CML experimental and material storage areas................................ 229

H-4 Room 101 walk-in hood contamination results, November 1989 to April 1990 ..................... 233

H-5 Air sample results summary for the CML Building 886 nuclear materials storage and 

experimental areas............................................................................................................... 238

H-6 Experiment periods, contamination incident dates, and associated records of airborne 

contamination results ........................................................................................................... 239
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ATTACHMENT H

CRITICAL MASS LABORATORY (continued)

H-7 HEU solution spills at CML................................................................................................... 243

H-8 Accumulated activities of dosimetrically significant MFAPs in UNH fuel at the CML............. 244

H-9 Maximum committed organ doses from inhalation of airborne MFAPs at CML..................... 245

LIST OF FIGURES

FIGURE TITLE PAGE

H-1 Weekly averages of removable alpha contamination in the Building 886 experimental 

and material storage areas, first quarter 1989...................................................................... 230

H-2 Average removable alpha contamination in the Building 886 experimental and material 

storage areas, second quarter 1989..................................................................................... 230

H-3 Weekly averages of removable alpha contamination in the Building 886 offices and 

control points (hallways and conference rooms), first quarter 1989...................................... 231

H-4 Average removable alpha contamination in the Building 886 office area, January to June 

1989..................................................................................................................................... 231

H-5 Floor plan of the assembly room (Room 101) showing the walk-in hood containing the 

vertical split table (V) and solution base (S) experimental locations ..................................... 234

H-6 Memorandum on beta/gamma penetrating and removable surface contamination 

measurements made upon discovering three sealed 90Sr sources in the Building 886 

material storage area ........................................................................................................... 235

H-7 Air sampler locations in the Building 886 cluster .................................................................. 237
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ATTACHMENT H

CRITICAL MASS LABORATORY (continued)

H.1 INTRODUCTION

This attachment assesses the potential internal exposure at the RFP CML. The CML operated at a 

maximum reactor power of 10 mW. A typical experiment duration of 1 hour was used to estimate the 

MFAP inventory built up over time in highly enriched uranium (HEU) solution fuel. 

H.2 PERSONNEL EXPOSURE CONCERNS

Fission products in irradiated fuels and activation products in both the fuel and containment materials 

are sources of external radiation dose to personnel using or working around the fuels, and they 

present an internal dose potential for personnel who might ingest or inhale them. Personnel 

dosimeters assigned to RFP radiation workers document the external exposures. Internal exposures 

might result (1) during operations from resuspension of contamination on surfaces, or (2) during 

facility demolition from airborne dust. CML staff was provided routine bioassay (urinalysis and wholebody counts) to detect intakes of plutonium, uranium, or americium, but MFAPs were not routinely 

monitored.

Surface contamination in CML experimental areas was extensive and predominantly due to spills of 

HEU in the form of UNH solution2 over the course of the facility’s history (Rothe 2005, pp. 447, 449, 

452, 454–458, 464, 467–471, 479, 486, 498, 500–502). Therefore, the MFAPs important from either 

an acute or chronic internal exposure perspective are those from CML criticality experiments involving 

HEU solution. MFAPs atom ratios on contaminated surfaces in the CML and resuspended in air from 

these surfaces are presumed to be the same as those in solution.

H.3 HISTORY OF URANYL NITRATE CRITICALITY EXPERIMENTS

Criticality experiments involving UNH solution were conducted from the beginning of CML operations 

until the last experiment in 1987. These experiments are documented in some detail in a published 

history of the facility (Rothe 2005, pp. 376–394). The 778 experiments performed at the CML with 

HEU solution or systems of HEU solution plus HEU metal (Oy) spheres took place in 10 campaigns 

between May 1967 and October 1987. 

H.4 REEVALUATED FISSION AND ACTIVATION PRODUCT LEVEL ESTIMATES

MFAP buildup in uranium solution experiments at the CML was calculated using ORIGEN-S, a 

program for calculating time-dependent concentrations of radionuclides that are simultaneously 

generated or depleted by processes such as fission, neutron absorption or transmutation, and 

radioactive decay. Initial assumptions about power levels and experiment durations were taken from 

a DOE document (DOE undated, p. 3) that stated that the experiments conducted in Building 886 

(housing the CML) generally involved power levels of no more than 10 mW for no more than 1 hour. 

It also stated that approximately half of the experiments in Building 886 actually achieved criticality, 

and only rarely were the radiation levels such that it was not possible to directly touch the fissile 

material and testing apparatus immediately after the experiments. 

RFP documents contain neutron flux and thermal power estimates by CML staff using analytical 

measurements made during or after criticality experiments. These calculations were performed for at 

 

2 Eleven spills involved solutions or dried salts from solution experiments, compared with two contamination events 

involving other solid fuels (one spill each of low-enriched uranium and plutonium powders).
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CRITICAL MASS LABORATORY (continued)

least three different EU fuel configurations: (1) HEU Oy spheres, (2) Oy spheres immersed in UNH 

solution; and (3) UNH only.

H.4.1 Thermal Power and Fission Rate Estimates in Bare Metal Experiments

Measurements of surface gamma fluence rates from a 7-cm-radius Oy sphere used in a 1967 1-hour 

criticality experiment led to an estimate of 1.9508 × 10-9 MW (1.9508 mW)3 thermal power, 

corresponding with 2.18 × 1011 total fissions4 (Rockwell 1967, p. 6).

H.4.2 Thermal Power and Fission Rate Estimates in Metal-Plus-Solution Experiments

CML staff made two estimates of total fissions in experiments involving Oy immersed in UNH. One 

1976 estimate used the residual gamma photon flux from the Oy + UNH to estimate an upper bound 

of 1.8 × 1018 total fissions over the 12 years of CML operation from 1965 to 19765 and an average 

power of 25 W (Rockwell 1976, pp. 2–4). However, the senior scientist who performed the 

measurements and derived the estimate noted that the gamma background from a new unirradiated 

Oy part was the same as the gamma flux from irradiated parts used for the analysis. This means that 

measurements made 6 years after the last criticality experiment were unable to detect additional 

gammas from previous experiments because of radioactive decay of the MFAPs. The original 

estimate did not take into account the intrinsic gamma emissions from unirradiated EU and was, 

therefore, invalid6 (Rockwell 1976, p. 2).

A second estimate of 3.44 × 1010 fissions in an average Oy + UNH experiment with a 20-minute run 

time was derived by estimating the number of neutrons produced (8.4 × 1010) using the current 

generated by a neutron detector associated with an April 1977 experiment (Rockwell 1977a, p. 3). 

This fission rate corresponds with a thermal power level of 0.92 mW.

H.4.3 Thermal Power and Fission Rate Estimates in Solution-Only Experiments

The CML staff’s most rigorous estimates of fission rates were derived for UNH (solution-only) fuel by 

measuring the rate of gamma decay in the fuel of Experiment 2-8-170. This experiment was 

configured as a 2 by 2 array of 8-in.-diameter tanks and was performed on May 4, 1977 (Rockwell 

1977b). The gamma count rate of a 2-mL aliquot withdrawn from the fuel after a 70.5-minute run was 

determined at 2- to 3-minute intervals using a “well crystal” calibrated to 137Cs photons. The total 

number of fissions in the reactor was calculated from the measured gamma emission rate (Rockwell 

1977c, p. 7):

 (H-1) 6 1.2 φ 1.9 10 τ photons/s/fission γ

− − ≅ ×

 

3 The conversion from MW to mW in the reference document incorporated a factor of 10 error, incorrectly concluding that 

the power level was 19.508 mW; the error is corrected here. 4 Using 1 W = 3.1 × 1010 fissions per second. 5 The estimate of total fissions is based on 27 days of continuous run time occurring in the middle of the 12-year period 

and corresponds with 7.7 × 1011 fissions per second. 6 The laboratory notebook entry reads: “All [gamma radiation is] due to fission 6 years ago. (No believable background 

available.) Note: Part no. 80 never used and yet had same background/kg as others. [Therefore] Feel fission too low 

to detect by this method.” 
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where:

φγ = reactor gamma photon emission rate at time τ

τ = days after shutdown

An initial evaluation7 of the data (Rockwell 1977b) arrived at a power estimate of 38 mW during the 

last 10 minutes of the experiment (when most of the corresponding 7.1 × 1011 fissions occurred), 

giving an average power of 5.4 mW for the 70.5-minute duration of the experiment.

A subsequent June 7, 1977, estimate from Experiment 2-8-170 employed more precise values for the 

aliquot and reactor volumes (Rockwell 1977c, p. 4). The estimate gave 25 mW at the end of the 

experiment, corresponding with 4.73 × 1011 fissions (or 3.6 mW average power). These results were 

within 2% of those obtained by the method of source multiplication (LANL 1977; Rockwell undated), 

using the equation below:

f st = μ (H-2)

where:

f = total number of fissions over time t

t = time

μ = average multiplication from the neutron detector response during the experiment

s = beginning neutron flux from the 252Cf seed source

On June 3, 1977, before the more precise June 7 calculation was completed, CML staff 

communicated officially to the U.S. Energy Research and Development Administration (ERDA). They 

provided an estimate of 8.8 × 1011 fissions and an average power of 6.7 mW for a typical 

(70.5-minute) experiment (Schuske 1977). There is no indication why the values reported to ERDA 

were almost 25% higher than results of the May 4, 1977 calculation.

H.4.4 Conclusions about Reactor Power and Fission Rates in Experiments

Table H-1 provides a summary of the different CML criticality experiments for which thermal power 

and fission rate were estimated from measurements.

CML staff described Experiment 2-8-170 to ERDA as having a higher-than-normal power level 

(Schuske 1977, p. 2). A thermal power estimate of 3.6 mW averaged over 70.5 minutes 

(corresponding with 4.73 × 1011 total fissions) was based on careful evaluation of measurements 

made on Experiment 2-8-170. This is considered the most precise estimate available for this 

relatively high-power experiment. CML staff, however, reported to ERDA an average thermal power 

of 6.7 mW over 70.5 minutes (8.8 × 1011 total fissions) based on the same experiment. The value 

reported to ERDA is considered the most appropriate value to represent the typical CML criticality 

experiments for the purpose of estimating doses from MFAPs. This value is less than the earlier 

assumption that power averaged 10 mW for experiments lasting an hour (1.1 × 1012 total fissions). 

However, the value is more favorable to claimants because it exceeds the most careful estimate of 

 

7 Reactor power was evaluated using the integral of Φ̇ over the operating time of the reactor, giving the gamma energy 

emission rate as a function of reactor power and time after shutdown.
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power for the experiment and because not all CML experiments achieved criticality (DOE undated, 

p. 3).

Table H-1. Average reactor power and fission rates derived from measurements at CML.a

Configuration Duration

Average power 

(mW)

Total 

fissions

Fission 

rate (s-1) Comment

Oy 60 min 1.9508 2.18E+11 6.06E+07 Based on exposure rate at the 

surface of an Oy sphere 

converted to γ photon flux.

Oy + UNH (a) 27 d 25,000 7.7E+11 3.30E+05 Estimate is invalid; measurement 

results were indistinguishable 

from background.

Oy + UNH (b) 20 min 0.92 3.44E+10 2.83E+07 From neutron detector current.

UNH (a) 70.5 min 5.4 7.1E+11 1.68E+08b Initial calculation on May 4, 1977, 

for Experiment 2-8-170

UNH (b) 70.5 min 6.7 8.8E+11 2.08E+08b Values reported to ERDA on June 

3, 1977, based on results of 

Experiment 2-8-170.

UNH (c) 70.5 min 3.6 4.73E+11 1.12E+08 Recalculated values from 

Experiment 2-8-170 on June 7, 

1977. Result is within 2% of that 

obtained from evaluation of 

source multiplication.

a. Sources: Oy from Rockwell (1967, p. 6); Oy + UNH (a) from Rockwell (1976); Oy + UNH (b) from Rockwell (1977a); 

UNH (a) from Rockwell (1977c, p. 7); UNH (b) from Schuske (1977); UNH (c) from Rockwell (1977c, p. 4).

b. These values are not reported in the original reference, but are calculated from total fissions and duration.

H.5 SURFACE CONTAMINATION 

H.5.1 Radiological Survey Results

Section H.9 lists SRDB Ref IDs that contain radiological survey results for Building 886 and 875 for 

January 1981 to December 1990. These files include results of daily removable contamination 

surveys in the office and experimental areas, and on equipment, as well as weekly penetrating dose 

measurements in the nuclear material storage and experimental areas of Building 886.

H.5.2 Maximum Reported Values

Based on the data, Table H-2 shows the maximum reported values of removable alpha surface 

contamination for each year.

Based on the data, Table H-3 shows maximum gamma and neutron penetrating dose rates.

H.5.3 Evaluation of Removable Contamination Survey Results

Contamination surveys were conducted daily at control points (hallways, doorways to offices), and 

weekly in offices within the unrestricted-access portion of Building 886. Surveys were conducted 

weekly in the experimental areas. Results of periodic contamination surveys in Building 875 and its 

air filtration plenums are also documented. Survey results for the office area have been found 

through December 20, 1990 (Rockwell 1990a, p. 778), and for the experimental and material storage 
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Table H-2. Maximum removable alpha contamination (dpm/100 cm2

).a,b

Year

Bldg. 886 

officesc

Bldg. 886 

experimental areac

Bldg. 886 

equipment

Bldg. 875

outside plenum

Bldg. 875

inside plenum

1981 30 366 156 12 11,196

1982 18 174 (3,930)e NS 12 14,346

1983f <20 576 NS NS NS

1984 <20 54,000 NS <20 15,660

1985 <20 174 NS <20 4,788

1986 18 186 (2,784) 3,954 12 5,394

1987 18 (24) 564 NS NS NS

1988 12 1,683 NS 12 15,240

1989g 18 18,707 NS NS NS

1990g 18 16,212 NS 6 NS

a. Source: See Section H.9. 

b. NS = not surveyed.

c. “Offices” includes control points (corridors and common areas) typically surveyed daily and individual offices surveyed 

weekly.

d. The experimental area (Rooms 101, 102, and 103) were surveyed weekly in 1981, but the frequency gradually 

diminished, apparently in response to decreased activities in these areas. Surveys were made approximately monthly 

by 1990.

e. Values in parentheses are singular outliers and not representative.

f. Data in 1983 are available only for November and December.

g. Elevated values of this order were observed in weekly surveys of the Room 101 walk-in hood from December 1989 

through April 1990, as discussed in the text.

Table H-3. Maximum dose rates in CML experimental and material storage areas.a,b

Year

Gammac

(mR/hr)

UNH Tank 

Room

Gammac

(mR/hr)

Solid Fuel 

Storage

Gammac

(mR/hr) 

Experiment 

Room

Neutrond

(mrem/hr) 

UNH Tank 

Room

Neutrond

(mrem/hr) 

Solid Fuel 

Storage

Neutrond

(mrem/hr) 

Experiment 

Room

1981 2.0 2.9 NS 0.3 12.9 NS

1982 2.0 2.5 3.7 0.2 14.7 5.4

1983 NS NS NS NS NS NS

1984 5.0 3.2 NS 0.2 0.6 NS

1985 2.0 2.8 NS 0.1 0.5 NS

1986 2.0 6.0 0.4 1.0 2.6 0.2

1987 3.0 2.0 3.0 1.7 2.0 3.4

1988 2.4 1.5 1.8 2.0 0.8 0.6

1989e 2.8 4.2 (250)f 0.0 1.0 1.8 (6.0) 0.6

1990 2.53 3.0 NS 0.3 0.5 NS

a. Source: See Section H.9. 

b. NS = not surveyed.

c. Gamma dose rate surveys were made at least weekly during the entire 10-year period.

d. Neutron dose rate surveys were made weekly, along with gamma surveys, until 1990 when the frequency was reduced 

to monthly.

e. Atypically elevated dose rate readings for July 6, 1989, are associated with a survey location within an area marked off 

inside the solid storage area.

f. Values in parentheses are singular outliers and not representative.

areas through April 1990 (Rockwell 1990a, p. 330). No average values were found, except in 1989 

documents, which contained a few graphs showing average contamination values in different areas 

for limited periods (Rockwell 1989a, pp. 3–6). These graphs are shown in Figures H-1 through H-4. 
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Figure H-1. Weekly averages of removable alpha contamination in 

the Building 886 experimental and material storage areas, first 

quarter 1989 (Rockwell 1989a, p. 3). 

Figure H-2: Average removable alpha contamination in the 

Building 886 experimental and material storage areas, second 

quarter 1989 (Rockwell 1989a, p. 5).
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Figure H-3. Weekly averages of removable alpha contamination in 

the Building 886 offices and control points (hallways and 

conference rooms), first quarter 1989 (Rockwell 1989a, p. 6). 

Figure H-4. Average removable alpha contamination in the 

Building 886 office area, January to June 1989 (Rockwell 1989a, 

p. 4). 

Table H-2 above, showing maximum values for the removable alpha survey results in both the office 

and experimental areas, demonstrates that radiological containment of the contaminated areas was 

quite effective. Excursions of removable contamination above the 20 dpm/100 cm2 DOE limit in “cold” 
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(office) areas is seen to occur only a few times (and only at low levels) in 1981 and once in 1984. 

Survey results in offices and hallways were typically well below 20 dpm/100 cm2

, particularly toward 

the end of the period from 1981 to 1990 (e.g., see Figures H-3 and H-4). The solid line in Figure H-4 

represents removable alpha contamination; the heavier dashed line represents the DOE limit for 

uranium.

Review of the survey documents indicates that excursions of removable alpha contamination above 

the Contamination Area limit in the experimental area (Room 101) and the material storage area 

(Rooms 102 and 103) were usually reduced below 1,000 dpm/100 cm2 within a few days, as noted 

below for entries in Table H-2. 

• The survey sheet showing 3,930 dpm/100 cm2 for removable contamination in Room 103 (the 

Mixing Room where uranium solution was housed) on September 10, 1982, for instance, has 

notation showing that the area was decontaminated to 12 dpm/100 cm2 on the same day 

(Rockwell 1982a, p. 23).

• A survey result of 54,000 dpm/100 cm2 on February 27, 1984 (the highest found in any of the 

documents), was in the walk-in hood of Room 101 (Rockwell 1984a, p. 13). A second survey 

sheet with the same date shows <20 dpm/100 cm2 for all survey locations in Rooms 101, 102 

and 103, and bears the handwritten note, “Retake of survey after [Redacted] and [Redacted] 

cleaned all of Rm 101.”

• The high value of 2,784 dpm/100 cm2 in 1986 is reported on a survey sheet dated June 17 

along with the handwritten notation, “Table Top 3954 d/m/100 cm2

; Rest of Table 

25.8 d/m/100 cm2

; deconned 6/18/86” (Rockwell 1986a, p. 8). The survey sheet for June 18, 

1986 indicates that all results in the controlled area were ≤24 dpm/100 cm2 (Rockwell 

1986a, p. 9).

• The 1988 high result of 1,683 dpm/100 cm2 was found just outside the walk-in hood of 

Room 101 on October 7, 1988 (Rockwell 1988a, p. 37). A similar elevated result 

(1,422 dpm/100 cm2

) was found inside the hood. Contamination at the location outside the 

hood had reduced to 60 dpm/100 cm2 a week later on October 14 (Rockwell 1988a, p. 38), but 

the elevated value inside the walk-in hood persisted. It had reduced to 126 dpm/100 cm2 in 

the October 21 survey report (Rockwell 1988a, p. 39).

Prompt decontamination practices were evident during most of the period from 1981 to 1990; 

however, elevated contamination levels in the Room 101 walk-in hood were recorded in December 

1989 and persisted through April 1990. The persistent contamination is reflected in the maximum 

recorded values of 1989 and 1990 in Table H-2 and is detailed in the weekly contamination survey 

results in Table H-4. The FBI raid of RFP in June 1989 resulted in the curtailment of plutonium 

operations. Afterward, the plant’s resources were redirected toward remediation of the issues 

resulting from the raid. According to A Technically Useful History of the Critical Mass Laboratory at 

Rocky Flats, “Manpower was not available to decontaminate Room 101 in the late 1980s because 

Maintenance personnel had all been dedicated to solving the plant’s larger problems” (Rothe 2005, 

p. 395). Handwritten notes on the survey sheets sometime indicate that full-face respirator protection 

was required in Room 101 when contamination levels exceeded the DOE limit (e.g., Rockwell 1989a, 

p. 109). Figure H-5 shows the floor plan for Room 101 where criticality experiments were performed; 

the walk-in hood is the enclosure left of center.
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Table H-4. Room 101 walk-in hood contamination results (dpm/100 cm2

), November 1989 to April 

1990.a

Date Location 20 Location 21 Location 22 Source Page

11/21/1989 27 642 NS Rockwell 1989a 92

11/30/1989 239 NS NS Rockwell 1989a 116

12/07/1989 897 12,579 18,707 Rockwell 1989a 160

12/14/1989 9 100 90 Rockwell 1989a 183

12/21/1989 72 51 5,235 Rockwell 1989a 196

01/18/1990 168 693 6,990 Rockwell 1990a 60

01/25/1990 84 672 6,963 Rockwell 1990a 81

02/01/1990 93 1,230 12,219 Rockwell 1990a 99

02/08/1990 105 1,290 13,452 Rockwell 1990a 117

02/15/1990 255 15,099 10,839 Rockwell 1990a 135

02/22/1990 717 12,369 9,978 Rockwell 1990a 156

03/01/1990 639 14,163 6,933 Rockwell 1990a 171

03/08/1990 616 14,193 6,987 Rockwell 1990a 183

03/15/1990 468 12,423 7,923 Rockwell 1990a 207

03/22/1990 1,035 16,212 5,403 Rockwell 1990a 234

03/29/1990 1,893 15,708 7,842 Rockwell 1990a 252

04/05/1990 2,748 12,951 8,967 Rockwell 1990a 273

04/19/1990 2,556 14,967 13,683 Rockwell 1990a 291

04/12/1990 1,572 13,593 10,353 Rockwell 1990a 300

04/26/1990 2,037 14,463 11,973 Rockwell 1990a 330

a. Locations 20, 21, and 22 correspond with the entry portal, Solution Base (S), and Vertical Split Table (V), respectively, 

as shown on the floor plan in Figure H-5. 

Air flow through Room 101 came from outside through the walk-in hood to the exhaust plenum in 

Building 875, then through high-efficiency particulate air filters before being exhausted to the outside 

air. The walk-in hood was, therefore, at a lower pressure than Room 101, drawing resuspended 

contamination to the exhaust plenum rather than the rest of Room 101. This engineered feature 

reduced exposure to Room 101 personnel and, along with workforce priorities redirected as a result of 

the FBI raid, probably explains why the hood was not decontaminated in late 1989 and 1990. The last 

CML criticality experiment concluded in October 1987 (Rothe 2005, p. 393); no routine work was 

performed in Room 101 after this date, further reducing the potential for personnel exposure as the

result of contamination in the walk-in hood in 1989 and 1990.

H.5.4 Discovery of Sealed Strontium-90 Sources

The file of survey results for June 1982 contains a copy of a note detailing radiological characteristics 

of three sealed 90Sr sources (Rockwell 1982b, p. 3). The note is reproduced as Figure H-6 below. It 

is the only documentation in hand indicating the presence of 90Sr at CML. There is no information 

about why the sources were at CML, how long they might have been there, or their disposition. 

Integrity of the source encapsulations was confirmed by removable contamination surveys of the 

sources and the cabinet in which they were found.

H.5.5 Conclusions About Surface Contamination

Surveys for removable alpha contamination at CML were conducted regularly. Excursions above the 

applicable DOE limits (20 dpm/100 cm2 in uncontrolled office areas and 1,000 dpm/100 cm2 in 

controlled-access experimental areas) were confined to discrete areas and were quickly 

decontaminated below the limits. Values above the limit in uncontrolled areas were only rarely
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Figure H-5. Floor plan of the assembly room 

(Room 101) showing the walk-in hood 

containing the vertical split table (V) and 

solution base (S) experimental locations 

(Rothe 2005, p. 120). 

observed. The largest measured value for removable contamination was 54,000 dpm/100 cm2

, found 

at sampling location 21 in the walk-in hood (Rockwell 1984a, pp. 13–14). This amount of 

contamination, if spread uniformly over the entire surface area8 of the controlled area, would be 

1,800 dpm/100 cm2

; it was decontaminated on the same day it was found. All other values for 

removable contamination correspond to <1,000 dpm/100 cm2

, if distributed uniformly. All were quickly 

decontaminated, except for that in the walk-in hood from November 1989 to April 1990, which was 

contained and ventilated during a time of little or no personnel activity in Room 101.

In light of the above information, the assumption that average removable contamination available for 

resuspension in the experimental and material storage areas (Rooms 101, 102, and 103) was equal 

to or less than the Contamination Area limit of 1,000 dpm/100 cm2 is favorable to claimants.

 

8 Dimensions of the walk-in hood (Rothe 2005, p. 129) in Room 101 were 3.0 m by 4.9 m. Sampling location 21 was in a 

part of the hood making up about half this surface area, 7.35 m2. Estimated floor areas for the individual rooms are 120 

m2 for Room 101 (Rothe 2005, p. 120); 40 m2 for Room 103 (Rothe 2005, p. 167); and 60 m2 for Rooms 102 and 108 

(connecting hallway) combined, assuming this area to be about half that of Room 101 (Rothe 2005, p. 110). The total 

estimated floor area is the sum of these values, 220 m2. Distributing the contamination at sampling location 21 over the 

entire surface area would result in an average contamination level reduced by the ratio 7.35/220.
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CRITICAL MASS LABORATORY (continued)

Figure H-6. Memorandum on beta/gamma penetrating 

and removable surface contamination measurements 

made upon discovering three sealed 90Sr sources in the 

Building 886 material storage area (Rockwell 1982b, 

p. 3). 

H.6 WORKPLACE AIR MONITORING

Bounds on internal dose from MFAPs were previously based on airborne concentrations calculated by 

applying a resuspension factor to surface contamination limits for the facility. This approach was 

based on interview comments from a former [redacted], who said that no routine air monitoring was 

performed at the CML before 1990, when this individual was assigned responsibility for Building 886 

(ORAUT 2015a, pp. 4, 7). Formal plantwide procedures describe a particulate air monitoring program 

during the period 1980 to 1989 for alpha emissions from uranium, plutonium, and americium at 

sampling locations selected by process knowledge or professional judgment. Additional documents 

indicate that these procedures appear to have been followed and that routine alpha air monitoring was 

performed at the CML during the period 1980 to 1989.

H.6.1 Workplace Air Monitoring Requirements

Formal procedures describe monitoring requirements and practices for routinely evaluating 

concentrations of alpha-emitting particulates (uranium, plutonium, and americium) in the RFP 

workplace air over the period 1979 through 1990 (RFP 1976–1996, pp. 19–136). The earliest known 

procedure (issued November 1979) and its subsequent revisions through December 1990 are listed 

below (RFP 1976–1996):

• Routine Air Sampling, HS-RM-4.1, November 13, 1979 (pp. 19–29);

• Routine Air Sampling, RMPM 4.1, August 2, 1982 (pp. 30–43);
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• Routine Air Sampling, RMPM 4.1, January 13, 1989, replaced August 2, 1982 (pp. 65–81);

• Routine Air Sampling, RMPM 4.1, June 1989, replaced January 13, 1982 (pp. 44–64);

• Routine Air Sampling, ROI 4.1, December 18, 1989, replaced June 15, 1989 (pp. 102–118); 

and

• Routine Air Sampling, ROI 4.1, Rev. 3, December 20, 1990, replaces December 19, 1989 

(pp. 119–136).

Review of the procedures shows consistent requirements for monthly calibration of gross alpha CAMs 

and selective alpha air monitors (SAAMs). CAM filters were collected and sent for analysis each 

weekday, except holidays, until at least January 1989, when RMPM 4.1 specified that the exchange 

frequency was to be determined by Operational Health Physics personnel (RFP 1976–1996, p. 67). 

Subsequent revisions assigned the responsibility for establishing sampling frequencies to Building 

123 Count Room personnel (p. 49), Operational Health Physics (p. 104), or to Radiological 

Engineering personnel (p. 121). When collected, filters were monitored with a handheld alpha survey 

instrument. Filters with excessive alpha activity were held for specified periods before analysis to 

allow decay of the short-lived radon and thoron progeny and a time-dependent Koval factor was 

applied to the analytical results to account for undecayed progeny (pp. 22–23, 36–37, 52, 72). Final 

results were reported as percentages of the RCG airborne limit for the material in question, or 

70 dpm/m3 for uranium9 (pp. 25, 39, 73). In June 1989, the airborne limit for uranium (in any form) 

was changed to 44 dpm/m3

, corresponding to the DOE Order 5480.11 (DOE 1988) derived air 

concentration (RFP 1976–1996, pp. 54, 104, 121).

H.6.2 Air Sampling Locations

Air particulate samplers were in Buildings 886 and 875, as marked on monthly air head calibration 

sheet maps (Rockwell 1987a to 1989c); Figure H-7 shows an example. The locations appear to be 

unchanged during the period 1980 to 1989. Building 875 housed the air exhaust plenums from the 

Building 886 experimental (Room 101) and material storage (Rooms 102 and 103) areas. The air 

head samplers were numbered RR-1 through RR-11 and fixed in the following locations:

• Building 886:

– RR-1 through RR-4, Room 103 (material storage);

– RR-5, Room 102 (material storage); and

– RR-6 and RR-7, Room 101 (experimental area).

• Building 875:

– RR-8 and RR-9, outside the plenum;

– RR-10, inside the first plenum; and

– RR-11, inside the tunnel from Building 886.

 

9 The RCG used by RFP in its air-sampling program corresponds with the maximum allowable (air) concentration used by 

AEC contractors in referring to the AEC’s “preferred level” of 50 μg/m3 for all uranium compounds on the basis of 

chemical toxicity. The level was also stated as 70 dpm/m3 for natural uranium (ORAUT 2006, p. 8).
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Figure H-7. Air sampler locations in the Building 886 cluster (Rockwell 1987–1988, p. 6). 

SAAM samplers were also in Rooms 101 to 103 of Building 886 and in the second plenum of 

Building 775. The detectors on these samplers were equipped with high-voltage discriminators that 

registered only alpha energies above a certain threshold, which provided a means to discriminate 

against lower-energy alpha emitters that would otherwise give false alarms. The filter material from 

the SAAMs was normally discarded without analysis unless the instrument had alarmed or the filter 

failed a handheld radiation monitor check (RFP 1976–1996, pp. 26, 35, 51, 70, 88, 105–106, 123).

Section H.9 includes files that contain air sampling results from 1981 to 1989 for samples labeled 

“875-A”, “875-B” and “875-C,” but their locations in Building 875 have not been identified.

CML personnel did not routinely access Building 875, and consideration of air monitoring results for 

this building is limited to those in the “RR” series, with known locations and direct correlation with daily 

samples from Building 886.

H.6.3 Air Sampling Results

Air sample results in 41 of 120 months have been found for the period of 1980 to 1989. Results from 

samplers RR-1 through RR-7 (Building 886) and RR-8 through RR-11 (Building 875) are available for 
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CRITICAL MASS LABORATORY (continued)

the period December 1980 to June 1989, although all but two of the results fall in the period from May 

1983 to November 1988. The data, summarized in Table H-5, indicate that procedural requirements 

for daily air monitoring appear to have been met at the CML. The number of analytical results per 

month for a particular sampler location is expected to be between 19 and 22, depending on the 

number of workdays. In the data, daily sample collection is demonstrated with certainty (results 

available for 19 or more days) in 15 of the 41 months (37% of the time for which records are available, 

but only 1% of the period from 1980 to 1989). Sample collection on at least an alternate-day average 

(results available for 10 to 18 days) is demonstrated with certainty during an additional 10 months. 

The availability of air sampling data does not seem to correlate with known spills or criticality 

experiments. Only one sampling day’s results were found for operations before 1983, but 207 

sampling days correspond with three operational periods after 1983. Four contamination incidents 

occurred from 1980 to 1989, but air monitoring results were only found for the period in which two 

incidents occurred closely together in 1987. Experiment periods, contamination incident dates, and 

associated records of airborne contamination results are summarized in Table H-6. 

Table H-5. Air sample results summary for the CML Building 886 nuclear materials storage and 

experimental areas.a,b 

Year Month Source Sample daysc

No. of days 

10%-99% RCGd

No. of days 

≥100% RCGe

1980 Dec Rockwell 1980 1 0 0

1983 May Rockwell 1983a 2 0 0

1983 Sep Rockwell 1983b 1 0 0

1983 Oct Rockwell 1983c 17 2 0

1984 Jul Rockwell 1984b 2 0 0

1984 Aug Rockwell 1984c 22 1 1

1984 Sep Rockwell 1984d 19 2 0

1984 Oct Rockwell 1984e 20 6 0

1984 Nov Rockwell 1984f 22 8 3d

1984 Dec Rockwell 1984g 15 5 0

1985 Jan Rockwell 1985a 19 4 0

1985 Feb Rockwell 1985b 19 8 2

1985 Mar Rockwell 1985c 22 1 0

1985 Apr Rockwell 1985d 21 0 0

1985 May Rockwell 1985e 22 3 0

1985 Jun Rockwell 1985f 22 0 0

1985 Jul Rockwell 1985g 23 2 0

1985 Aug Rockwell 1985h 20 2 0

1985 Sep Rockwell 1985i 15 0 0

1985 Oct Rockwell 1985j 9 0 0

1985 Nov Rockwell 1985k 19 1 0

1985 Dec Rockwell 1985l 15 0 0

1986 Jan Rockwell 1986b 19 0 0

1986 Feb Rockwell 1986c 1 0 0

1986 Jun Rockwell 1986d 4 0 0

1986 Jul Rockwell 1986e 14 0 0

1986 Nov Rockwell 1986f 3 0 0

1986 Dec Rockwell 1986g 9 0 0

1987 Jan Rockwell 1987a 15 1 0

1987 Feb Rockwell 1987b 22 0 0

1987 Mar Rockwell 1987c 8 0 0

1987 May Rockwell 1987d 6 0 0

1987 Jul Rockwell 1987e 3 0 0
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CRITICAL MASS LABORATORY (continued)

Year Month Source Sample daysc

No. of days 

10%-99% RCGd

No. of days 

≥100% RCGe

1987 Aug Rockwell 1987f 13 0 0

1987 Sep Rockwell 1987g 7 0 0

1987 Dec Rockwell 1987h 11 0 0

1988 Jan Rockwell 1988c 18 0 0

1988 Feb Rockwell 1988d 15 0 0

1988 Apr Rockwell 1988e 5 0 0

1988 Nov Rockwell 1988f, 1988g 5 0 0

1989 Jun Rockwell 1989c 1 0 0

Totals N/A N/A 526 46 6

a. N/A = not applicable.

b. Documents include results from only portions of years 1980 and 1983 to 1989, as shown in the table.

c. The number of days in the specified month for which air monitoring data indicate that samples were collected. 

Procedures required that sample filters be changed and analyzed daily, except over weekends and holidays.

d. The number of days for which the specified fraction of the uranium RCG of 70 dpm/m3 alpha was met or exceeded.

e. Notations on the air results indicate that respirators were worn.

Table H-6. Experiment periods, contamination incident dates, and associated records of airborne 

contamination results. 

Experimental campaignsa Contamination incident datesb Airborne monitoring days captured

None 11/25/1980 0

02/1978–09/1981 None 1

Summer 1982–12/20/1982 None 0

05/1983–09/1984 None 64

None 07/07/1984–07/20/1984 0

07/1985–08/1986 None 139

None 02/14/1987 and 03/13/1987 30

04/1986–10/1987 None 104

a. Source: Rothe (2005, pp. 389–393).

b. Source: Rothe (2005, pp. 486–487, 498, 500–501).

Building 886/875 results appear in the same reports as those for Building 865, which housed the 

Metal Research and Development Laboratory10 and which processed nonplutonium metals including 

DU (DOE 2011b, p. 2). Reports were reviewed and initialed, and instructions on the report sheet 

specify that a copy was to be sent “TO RADIATION MONT 881” (to Radiation Monitoring, 

Building 881).

Individual sample results are occasionally lined out in the reports, with or without explanation. There 

are multiple instances when results are lined out with the notation “W/C” or “Wrong Color.” The 

meaning of this notation is not clear, and these samples are not included in evaluating the results.

A total of 526 days of results were evaluated against the RCG of 70 dpm/m3

. One or more results 

from the seven samplers in Building 886 exceeded the RCG on 6 days; results between 10% and 

100% of the RCG were recorded on 46 additional days. All results were below 10% of the RCG for 

the remaining 473 days. Respirators were worn on 3 of the 6 days in which the RCG was exceeded 

(Rockwell 1984b, pp. 33, 36, 52). Results on the other 3 days in which the RCG was exceeded were: 

522.02% RCG (Rockwell 1984c, p. 5), 111.72% RCG (Rockwell 1985c, p. 37), and 117.76% RCG 

(Rockwell 1985c, p. 41).

 

10 Each day’s results for Building 886 are labeled RR-1 through RR-7; those for Building 875 are labeled RR-8 through 

RR-11; Building 865 results begin with a UU- designator.
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CRITICAL MASS LABORATORY (continued)

Results from Building 875, which housed the effluent air exhaust plenums from the Building 886 

experiment room, exceeded the RCG on only one occasion11 during the same 526 days; results 

between 10% and 100% of the RCG were recorded on 2 days.

H.6.4 Conclusions About Workplace Air Monitoring

A robust and well-defined workplace air monitoring program for alpha-emitting radioisotopes was 

required by RFP procedures on a continuing basis during the period from 1980 to 1989. In all air 

monitoring records, air particulate samples from the CML, Building 886, and its air exhaust filtration 

plenums in Building 875 were routinely analyzed and reported along with those from the Metal 

Research and Development Laboratory, Building 865. Sampling results were evaluated for uranium 

alpha emissions and reported as a percentage of the RCG airborne limit for uranium, 70 dpm/m3

. The 

analytical reports were reviewed and initialed, and were to be sent to Radiation Monitoring in Building 

881. Excursions in excess of the RCG were uncommon, occurring in Building 886 on 6 of 526 days 

for which monitoring results were found and in Building 875 only once.

A bounding value for activity concentrations in breathing air can be calculated as the weighted 

average C̅of air results using recorded values for three results in excess of the RCG with no 

indication that respirators were worn, and by making the assumptions that results were 70 dpm/m3 for 

46 results between 10% and 100% of RCG and 7 dpm/m3 for the remaining 477 recorded samples 

(which are favorable to the claimant), as shown in the equation below:

(H-3)

( ) ( ) ( ) ( )

3

5.22 1.12 1.18 46 1.0 477 0.1

RCG-d

526d

 13.5 dpm/m

C

  +++ +   =

=

Results for the 526 days of monitoring data found for the period from 1980 to 1989 are assumed to be 

representative of data that have not been found for that period for the following reasons:

• The same plant health physics procedures requiring air monitoring were in effect continuously 

(with revisions) for the entire period,

• The assumption is made that daily samples were collected and analyzed in accordance with 

plant procedures over the entire period, and

• The results are probably similar because CML operations were similar over the period.

These data were available for routine review by health physics personnel, who also had access to 

information about operations for making personnel monitoring decisions. It is therefore unlikely that 

an unrecorded intake of alpha-contaminated airborne particulates occurred during this period because 

of a lack of relevant air monitoring data.

H.7 ASSESSMENT OF UNMONITORED RADIATION DOSE

Radiation dose from intake of MFAPs at CML could have occurred during cleanup of numerous fuel 

spills, predominantly from enriched UNH solution, or from inhalation of dried, resuspended 

 

11 The sample designation was RR-12 for the one result exceeding the RCG, without explanation of the location or 

purpose for the sample. A location in Building 875 is assumed.
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CRITICAL MASS LABORATORY (continued)

contamination deposited on surfaces as the result of these spills (Rothe 2005, pp. 447–449, 452–462, 

464–473, 486–487, 498, 500–501). RFP workers (including those assigned to the CML) with the 

potential for receiving intakes of plutonium, americium, or uranium were monitored by periodic 

urinalysis and body counts (NIOSH 2006, p. 30). However, there is no indication that confirmatory 

bioassays were performed for workers involved in cleanup of any of the accidental UNH spills. 

MFAPs, which decay primarily by beta/gamma emission, are not likely in any case to have been 

detected by bioassay intended to detect alpha particles from uranium or transuranic radionuclides.

Maximum MFAPs internal doses to CML workers were estimated by modeling a representative UNH 

experiment and calculating the MFAPs inventory based on the historical record of CML experiments 

with UNH, and on the average thermal power and duration of CML UNH criticality experiments. 

Intakes of resuspended UNH contamination with the same MFAPs-to-uranium atom ratio as the fuel 

were estimated from the weighted average of air monitoring results in the experimental and materials 

storage areas of the CML. Doses were calculated by applying ICRP Publication 68 (ICRP 1995) 

DCFs for three solubility categories of dosimetrically significant radionuclides, using the method 

described in ORAUT-OTIB-0054, Fission and Activation Product Assignment for Internal DoseRelated Gross Beta and Gross Gamma Analyses (ORAUT 2015b).

H.7.1 Calculation of Fission and Activation Product Content of Uranyl Nitrate Solution

A series of experiments was performed at the CML during the mid-1970s to determine the critical 

height of UNH in suspended cylindrical tanks. Experiments were performed using different uranium 

concentrations, different tanks, and with and without neutron reflectors. The same uranium 

enrichment was used in all experiments; only the concentration was varied.

One of the unreflected suspended tank experiments from the mid-1970s was chosen to represent 

UNH experiments performed over the CML’s history. There were 10 unreflected suspended tank 

experiments in all. Two of them used a stainless-steel tank and the rest used an aluminum tank.

One of the experiments that used the stainless-steel tank was selected so that the calculated MFAP 

content would include iron activation products. The tank had an inside diameter of 27.92 cm and an 

inside height of 41.6 cm. For the selected experiment, the tank was filled with UNH with a uranium 

enrichment of 93.172 weight percent 235U at a concentration of 145.68 g/L. The other unreflected 

stainless-steel tank experiment used UNH with the same enrichment, but at a concentration of 

346.73 g/L. The lower concentration was selected because it represents the middle of the range of 

concentrations used across the 10 unreflected tank experiments. The critical solution height for the 

selected experiment was found to be 31.20 cm. The selected experiment is documented as Case 

Number 1 in “Minimally Reflected Cylinders of Highly Enriched Solutions of Uranyl Nitrate,” HEU-SOLTHERM-001, from the International Handbook of Evaluated Criticality Safety Benchmark Experiments

(Palmer 2004).

The MFAP composition for the selected experiment was calculated using SCALE. SCALE is a 

modular system of computer codes for nuclear- and radiological engineering-related analyses from 

ORNL. SCALE’s TRITON module was used to develop a case-specific cross-section library for the 

selected suspended tank experiment. The library was then used by the ORIGEN-S code to determine 

the time-dependent MFAP content of the UNH and the stainless-steel tank over the CML’s history.

The TRITON module performs neutron transport and depletion calculations for irradiated nuclear 

fuels. Neutron transport is performed using either discrete ordinates methods (via the NEWT code) or 

via Monte Carlo calculations (using the KENO-V or KENO-VI codes). TRITON’s capabilities include 
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ATTACHMENT H

CRITICAL MASS LABORATORY (continued)

creating case-specific cross-section library files that can be used by other elements of the SCALE 

code system, notably the ORIGEN-S code. Among the numerous capabilities of ORIGEN-S is 

calculation of the inventory of activation products, actinides, and fission products in a composition as 

a function of time and burnup (as applicable).

TRITON was used to model the stainless-steel tank and UNH solution geometry of the selected 

suspended tank experiment as a system of cylinders representing the tank bottom, tank walls, and the 

UNH volume. The tank and solution heights were truncated to the critical solution height of 31.20 cm. 

Neutron transport was performed via the KENO-VI Monte Carlo code.

The case-specific cross-section library created by TRITON was subsequently used in an ORIGEN-S 

calculation to determine the MFAP content of the solution and the tank volume at the end of the tenth 

campaign and thereafter. This reflects a modeling assumption that the same tank and solution were 

used for all 778 experiments involving HEU solutions at the CML over its operating history. The 

composition used for the ORIGEN-S calculations was a homogenized mixture of the HEU solution, 

the 304 stainless-steel tank walls and bottom, and the associated impurities. The composition was 

irradiated and decayed in the ORIGEN-S case using the history in Table H-6 above. The first 

campaign was represented by a 38-hour irradiation followed by 32 days of decay, and so on, through 

the final 19-hour irradiation representing the tenth campaign.

Each experiment was assumed to have lasted for 70.5 minutes and to have produced an average 

thermal power12 of 6.7 mW (the average power and duration reported to ERDA in 1977) (Schuske 

1977, p. 6).13

The ORIGEN-S calculation produced time-dependent MFAP inventories, with radioactive decay 

between each of the 10 campaigns, for the HEU solution defined in the selected benchmark case, and 

using the timeline for all uranium solution experiments conducted over the CML’s history. Radioactive 

decay corrections were also applied at the end of the final experiment, using a decay period of 

180 days from those specified in ORAUT-OTIB-0054 (ORAUT 2015b).

H.7.2 Calculation of Inhalation Intakes and Committed Organ Doses

Solution spills resulting in surface (floor) contamination occurred throughout CML operations, as 

shown in Table H-7. Most spills occurred in the late 1960s, but there were several spills during the 

1980s. Organ dose calculations were performed after applying a decay interval of only 180 days to 

the ending MFAP inventory for the HEU solution. This is favorable to the claimant given that the 

calculation of the MFAP inventory represents the entire operating history of the facility (i.e., includes 

the period after the last major spill).

 

12 The power in this context is an average value representing a given number of fissions over a given interval of time 

because nuclear criticality experiments are not steady state and are not typically considered in terms of a power level. 

An average power was used for modeling purposes to account for change in the composition of the fissile solution (i.e., 

for depletion effects) and to compute the ingrowth of fission products as the solution was used. In reality, most of the 

fissions would have occurred near the end of a given experiment when the system was alternately placed in slightly 

subcritical and slightly supercritical states. 13 The 6.7-mW value is used, rather than the more precise 3.6 mW for 70.5 minutes estimated by CML staff after the 

communication with ERDA. There is no indication why 6.7 mW was reported instead of a lower value of 5.4 mW initially 

calculated by CML staff before refining their estimate. The higher value was adopted as more favorable for claimants.
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CRITICAL MASS LABORATORY (continued)

Table H-7. HEU solution spills at CML.a

 

Date

Volume 

(L)

Uranium mass

(kg) Contaminated area Page(s)

07/02/1965 0.5 0.225 22 m2 (floor) 447

07/14/1965 Unknown Unknown 10 m2 (floor) 448–449

07/22/1965 Unknown Unknown Small amount (floor) 449

11/30/1967 Unknown 9 (Inside large duct, filter housing, vent line)b 452–462

02/16/1968 Unknown 1.14c (Floor and cable trenches) 464-467

05/11/1968 0.06 Unknown (Worker’s knee) 467

05/09/1969 150.1 16.1 20 m2

(mixing room floor)d 467–473

11/25/1980 7 2.66 (Assembly room hood) 486–487

07/07–20/1984 Unknown Unknown (Walk-in hood)e 498

02/14/1987 Unknown Unknown (Personnel, facility and fixtures)f 500–501

03/13/1987 Unknown Unknown (Personnel, facility and fixtures)f 501

a. Source: Rothe (2005).

b. Although contamination was confined to ducts and a filter housing, cleanup of this incident resulted in a blowback of 

dried salts, resulting in facial contamination of a staff member.

c. Two conflicting accounts refer to this value as either the solution mass or uranium mass.

d. Standing HEU solution covering the floor was cleaned up by a staff member using a critically safe vacuum and wearing 

plastic booties and a half-face respirator.

e. Potentially contaminated workers repairing a [redacted] were required to evacuate when a criticality alarm was triggered 

by an electrician.

f. Two essentially identical events resulted in personnel, fixture, and facility contamination by resuspended HEU UNH salts 

accidentally knocked from the surface of a large reactivity shim.

ORAUT-OTIB-0054 (ORAUT 2015b) describes a method to reduce the large number of MFAP 

isotopes in an ORIGEN-S result to a set of 36 dosimetrically significant nuclides. The same 36 

nuclides were considered in the inhalation intake and committed organ dose calculations for the CML. 

Intakes were computed using the 180-day activity values for the HEU solution shown in Table H-8, 

corrected for the average airborne concentration level, and assuming an intake period of 4,000 hours 

(2 working years) at a breathing rate of 1.2 m3

/hr. The 2-year intake period was selected to be 

consistent with ORAUT-OTIB-0054, which provides a basis for assigning internal dose from unknown 

inhalation of MFAP mixtures. Radioactive decay of the isotopic mixture over the 2-year intake period 

was not considered, which is favorable to the claimant.

Note that the inventory in the UNH fuel was accumulated without radioactive decay until the final 

experiment, after which the activities of individual isotopes were decay-corrected for 180 days to yield 

these values.

The evaluation of air monitoring data (Section H.6), determined that a weighted average concentration 

of 13.5 dpm/m3 for airborne alpha activity was favorable to the claimant. Inhalation intakes were 

computed by assuming that the airborne alpha activity consisted entirely of HEU having a specific 

activity of 70 μCi/g. Therefore, the airborne mass concentration was 8.7 × 10-8 g/m3 and the total 

intake over the 4,000-hour period was 4.2 × 10-4 g.

The UNH fuel, as modeled, contained 2,782.8 g HEU. The ratio of total inhalation intake over the 

2-year intake period and HEU solution mass gives an intake fraction f = 1.5 × 10-7. The inhalation 

intake I for the 36 dosimetrically significant nuclides for the 2-year period was then determined as the 

product of the ORIGEN-S result for each nuclide after 180 days of decay (activity A) and the intake 

fraction (i.e., I = fA).
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ATTACHMENT H

CRITICAL MASS LABORATORY (continued)

Table H-8. Accumulated activities of dosimetrically 

significant MFAPs in UNH fuel at the CML. 

Isotope Activity (Bq)

Mn-54 1.800E+02

Fe-55 1.664E+03

Co-58 1.359E+02

Co-60 3.077E+00

Sr-89 1.287E+04

Sr-90 2.506E+04

Y-90 2.507E+04

Y-91 1.984E+04

Zr-95 2.734E+04

Nb-95 5.025E+04

Mo-99 5.206E-14

Ru-103 4.732E+03

Ru-106 3.889E+03

Cd-113m 1.658E+00

Cd-115m 9.955E-01

Sb-125 3.236E+02

Te-129m 1.051E+02

Te-132 3.937E-11

I-131 8.936E-02

I-132 4.055E-11

Cs-134 7.004E-02

Cs-136 4.318E-02

Cs-137 2.567E+04

Ba-140 3.719E+01

La-140 4.281E+01

Ce-141 5.509E+03

Ce-144 5.417E+04

Pr-143 6.690E+01

Pr-144 5.417E+04

Nd-147 3.204E+00

Pm-147 2.171E+04

Pm-148m 0.000E+00

Sm-151 6.867E+02

Eu-154 1.691E-03

Eu-155 3.225E+02

Ta-182 0.000E+00

Committed organ doses were computed as the sum of products of the 180-day intakes, computed for 

each of the 36 dosimetrically significant nuclides and their corresponding inhalation DCFs from ICRP 

Publication 68 (ICRP 1995):

(H-4)

36 36

1 1

ii ii

i i

H(50) I DCF f A DCF = == = ∑ ∑

The DCFs were partitioned into three solubility categories to account for different absorption types, 

using the method described in ORAUT-OTIB-0054 (ORAUT 2015b). Three committed dose values 

were therefore computed for each of the 25 individual organs defined in ICRP Publication 68 (ICRP 
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ATTACHMENT H

CRITICAL MASS LABORATORY (continued)

1995), corresponding to soluble, moderately soluble, and insoluble materials. Table H-9 shows the 

maximum committed organ dose values for the three solubility categories.

Table H-9. Maximum committed organ doses from 

inhalation of airborne MFAPs at CML. 

Absorption type

H(50)

(Sv) Tissue

Type F 2.5 × 10-9 Bone surface

Type M 2.3 × 10-9 Bone surface

Type S 2.4 × 10-9 Lung

These values are much less than previously calculated committed doses of 3.7 × 10-7 Sv (soluble, 

bone surface), 4.0 × 10-7 Sv (moderately soluble, lung), and 6.1×10-7 Sv (insoluble, lung). Earlier 

committed doses were calculated using airborne concentrations derived by applying a resuspension 

factor to the DOE limit on removable surface contamination (which is favorable to claimants). The 

orders-of-magnitude difference in the two sets of dose values is due to a previous miscalculation in 

converting from 100 cm2 to m2

. The corrected calculation uses a resuspension factor of 1.5 × 10-4 m-1

and 2,000 dpm/100 cm2 removable alpha. This calculation is applied over the entire 220 m2 footprint 

of the CML experimental and material storage areas, giving an estimated air concentration of 

30 dpm/m3

. The 13.5-dpm/m3 weighted average alpha air concentration from routine monitoring 

results used in calculations for this analysis further reduces the dose estimate from that obtained 

previously.

Previous estimates of MFAP inventories assumed a thermal power of 10 mW for 60 minutes, 

compared with the lesser (but still favorable to claimants) value of 6.7 mW for 70.5 minutes used in 

this analysis to describe conditions for the typical CML criticality experiment.

H.8 SUMMARY AND CONCLUSIONS

An evaluation of the unmonitored personnel dose, air monitoring data. and reactor performance 

estimates for the CML has resulted in estimates on the order of a few nanosieverts for maximum 

organ doses due to inhalation of resuspended contamination containing MFAPs. The greatest 

contributor to the large reduction in estimated doses is a correction in the calculation of a conversion 

factor. Lesser contributors to the reduction are lower estimates of reactor power in a typical criticality 

experiment at CML, and a lower value for respirable alpha air concentrations based on routine air 

monitoring results.

Based on this modeling, no significant personnel dose to RFP workers or contractors resulted from 

the generation of MFAPs in the UNH fuel or resuspended contamination from fuel spills as a result of 

criticality experiments at CML over its lifetime.

H.9 DATA SOURCES

The following are the complete lists of survey and sampling results that were reviewed as part of this 

evaluation:

• Radiological survey results. 156807, 156808, 156809, 156810, 156811, 156812, 156813, 

156814, 156815, 156816, 156817, 156818, 152233, 152245, 152247, 152248, 152250, 

152251, 152253, 152255, 152257, 152259, 152261, 152262, 152263, 152264, 152266, 

152267, 152268, 152269, 152270, 152276, 152281, 152288, 152292, 152295, 156820, 

152271, 152272, 156822, 156824, 156825, 156826, 159203, 159204, 159205, 159207, 
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CRITICAL MASS LABORATORY (continued)

159210, 152229, 152273, 152277, 152280, 152284, 152286, 152289, 152293, 152299, 

152301, 152302, 152304, 152318, 152323, 152325, 152331, 152335, 152340, 152342, 

152343, 152344, 152346, 157168, 157170, 157172, 157173, 152298, 152306, 152308, 

152310, 152312, 152314, 152317, 152320, 152326, 152351, 152353, 152355, 152327, 

152328, 152329, 152330, 152332, 152341, 152348, 152357, 152359, 152360, 152362, 

152500, 152502, 152507, 152508, 152509, 152511, 152515, 152516.

• Air sampling results. 159122, 159124, 159125, 159126, 159127, 159128, 159129, 159130, 

159131, 159132, 159133, 159135, 159137, 159138, 159136, 159154, 156834, 159155, 

159158, 159160, 159162, 156857, 156858, 156859, 156860, 156861, 156862, 156864, 

156875, 156877, 156879, 159139, 156863, 156867, 156876, 156878, 160323, 159142, 

159143, 159144. 159145. 159146, 159147, 159148, 156881, 156883, 156884, 156885, 

156886, 156888, 156890, 156891, 156892, 156893, 156894, 156895, 156897, 156880, 

156882, 156887, 156889, 160325, 156896, 159140, 159149, 159123, 159150, 159151, 

159152, 159153, 159156, 159157, 156898, 156899, 156901, 156903, 156905, 156906, 

159141, 156900, 156902, 156907, 159159, 159161, 159163.
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ATTACHMENT I 

FEDERAL BUREAU OF INVESTIGATION RAID

TABLE OF CONTENTS

SECTION TITLE PAGE

I.1 Introduction .......................................................................................................................... 248

I.2 Interviews............................................................................................................................. 248

I.3 An Insider’s View of Rocky Flats: Urban Myths Debunked .................................................. 249

I.4 Followup Research: Availability and Accessibility of Relevant Documents .......................... 250

I.4.1 Additional Colorado Visit for Data Capture and Interviews........................................ 251

I.4.2 Site Visit and Data Capture Interviews after Denver Visits........................................ 254

I.4.2.1 Data Falsification Interviewee..................................................................... 254

I.4.2.2 FBI Raid Interview and Followup Information Reviews............................... 254

I.4.3 Review of Petitioner-Identified Technical Safety Appraisal Issues ............................ 257

I.4.4 Review of Petitioner-Identified Data Falsification Issues Report ............................... 258

I.4.5 Additional Information from RFP Manager Personal Notebooks ............................... 259

I.5 Assessment of Available Personnel Radiological Monitoring Data ....................................... 259

I.6 General Summary and Conclusions..................................................................................... 260

I.7 Examples of Post-Raid Procedures and Manuals ................................................................ 263

LIST OF TABLES

TABLE TITLE PAGE

I-1 Available internal sample data, 1980 to 1995....................................................................... 260

LIST OF FIGURES

FIGURE TITLE PAGE

I-1 Percent of workers badged for dosimetry ............................................................................. 260
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ATTACHMENT I

FEDERAL BUREAU OF INVESTIGATION RAID (continued)

I.1 INTRODUCTION

This evaluation originated from the followup review of the SEC-00192 RFP evaluation report (NIOSH 

2013a) by the RFP ABRWH Working Group. The SEC-00192 petitioner initially identified issues 

relating to allegations made by a former RFP worker in interviews on October 24 and 25, 1989, by the 

EPA National Enforcement Investigation Center’s Office of Criminal Investigations and the FBI. 

These interviews resulted from a telephone call by the interviewee to the FBI Rocky Flats Hotline on 

June 16, 1989, alleging safety violations and manipulation of laboratory samples at RFP. A redacted 

transcript of the interviews was provided by the EPA Office of Criminal Investigations (EPA 1989a) 

and reviewed for this evaluation. Specifically, the EPA and FBI interviews were technically reviewed 

to assess the allegations and their relevance to potential data falsification and data invalidation in 

Building 123 in the context of the technical basis for dose reconstructions under EEOICPA. 

I.2 INTERVIEWS

Most of the information and incidents described by the FBI interviewee do not provide sufficient detail 

to support a followup investigation of the claims. Four individuals with potential related knowledge or 

information that might have provided insight into the statements offered by the FBI interviewee were 

identified and interviewed. A summary of the interviews is presented below.

• A former RFP [redacted] who worked from [redacted] to [redacted] managing both [redacted] 

(ORAUT 2013h) was interviewed. He returned as a worker from [redacted] to [redacted] 

working in [redacted]. In [redacted] to [redacted], he also worked with the [redacted]as 

manager of their [redacted]. As such, his RFP experience was related to the period directly 

after the raid (he had no information or experience relating to the period before the raid). 

These observations have relevance if one assumes that the 1989 protocols in place in the FBI 

investigation timeframe were essentially the same as those in place when he began work in 

1990, which is supported by Building 771 Laboratory sample-handling procedure reviews 

before and after the period around the 1989 raid (Fraser 1996). During the interview, the 

following observations were made (summarized):

– Environmental monitoring and personnel dosimetry were separate programs, although 

their respective samples were analyzed in the same low-level onsite laboratory. 

Around 1997 to 1998, the onsite laboratory was shut down and everything was 

contracted out. Turnaround times on samples was sometimes a problem, especially 

for plutonium, but not as much for tritium. While the interviewee was there, there was 

no routine tritium monitoring program, just some pre- and post-job tritium analyses. 

There were no significant tritium intakes during his employment.

– Based on his RFP experience, there is no specific link between environmental and 

bioassay results; the same numbers might have a different significance in 

environmental versus bioassay samples.

– RFP had a state-of-the-art program for dealing with compromised personal protective 

equipment. They had CAM alarms, nasal swipes, and bioassay. Bioassay would be 

done immediately if there was a suspected exposure, not necessarily at the end of the 

work shift. RFP also had a wound counting program.

– In bioassay analysis, the RFP laboratory staff used hoods. There was a complete 

industrial hygiene staff. The interviewee is sure the airflow was tested and is not aware 
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of any injuries or complaints about hoods. Strong acids were used in bioassay for fecal 

samples, but the interviewee recalled no incidents.

• An individual serving as a RFP [redacted] from [redacted] to [redacted] and an RFP [redacted] 

from [redacted] to [redacted] was interviewed (ORAUT 2013i). Based on the employment 

period, this individual also only had RFP experience outside of the period of the FBI raid. The 

interviewee had no information or experience about the time before the raid and had no 

specific information or feedback pertinent to this review of the raid or data falsification at RFP.

• A former RFP worker indicated that the raid involved specific people; therefore, information 

was provided only on a need-to-know basis (i.e., those not specifically involved in the 

investigation received no information about what was going on). Because the raid was related 

to environmental issues (as opposed to occupational radiological issues), there was no 

involvement from the bioassay program perspective and no radiological program changes 

were made as a result of the raid. This interviewee was not informed of any aspects of that 

raid, but indicated that the department did not know the raid had happened until it was in the 

news. The interviewee also indicated that, at the time of the interview, no information had 

been presented about any aspect of the raid (ORAUT 2013j).

• A second former RFP worker confirmed that the raid involved specific people and only those 

who were involved received information about what was going on. The interview confirmed 

that the raid was related to environmental issues and not occupational radiological issues. 

Because there was no involvement from the bioassay program perspective, there were no 

radiological program changes made as a result of the raid (ORAUT 2013k).

Based on the information from the interviewees, the focus of the assessment and the raid was very 

specific to environmental impacts and monitoring. Because the personnel radiological monitoring 

program was not involved in the raid or associated assessment, there is no direct translation of the 

identified environmental findings and deficiencies into findings or deficiencies in the worker monitoring 

program. Further, no such formal allegations were made. Personal monitoring is the primary focus of 

individual dose reconstruction under EEOICPA.

I.3 AN INSIDER’S VIEW OF ROCKY FLATS: URBAN MYTHS DEBUNKED

An Insider’s View of Rocky Flats: Urban Myths Debunked by a former RFP worker was reviewed for 

this attachment (Hobbs and Warden 2010). The author implies (and most articles accessed via the 

Internet seem to agree) that the FBI raid on RFP found no issues with worker protection or the worker 

monitoring program. The only violation cited for RFP was an environmental release.

An audit was performed by a DOE Special Assessment Environmental Team that focused on 

environmental issues, finding 95 deficiencies of varying types (DOE 1989a). The following two 

deficiencies mentioned the Building 123 laboratory:

• Under Radiation. Quality assurance (QA) and quality control (QC) practices for radiochemistry 

analyses in the Building 123 Health, Safety, and Environment Laboratory do not conform to 

generally accepted practices. The description of the finding further defines that the analyses 

were specific to environmental monitoring at RFP.
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• Under Quality Assurance. QA and QC practices at Building 123 Health, Safety, and 

Environment Laboratory for environmental analyses are not adequate to document validity of 

data.

Based on the information in the DOE Special Assessment report, the focus on the assessment was 

very specific to environmental impacts and monitoring. Although the report discusses data QA and 

validation issues with the analytical laboratories in Building 881 and Building 123, there were no 

situations identified in which falsification or invalidation of data would affect the ability to perform dose 

reconstruction under EEOICPA (DOE 1989a, Sections 7 and 8).

I.4 FOLLOWUP RESEARCH: AVAILABILITY AND ACCESSIBILITY OF RELEVANT 

DOCUMENTS

The initial documentation that related to the raid and subsequent litigation includes:

• A report from a DOE Environmental Special Assessment Team (one of four assessment 

teams that also included Management and Operations, Safety, and Legal Matters) (DOE 

1989a). These teams were mobilized by the Secretary of Energy to perform a separate 

evaluation in parallel with the FBI investigation in order to provide the department with an 

independent assessment of RFP at the time of the raid.

• A 1995 symposium presentation titled Are You Prepared To Survive an FBI Raid At Your 

Facility? This specifically discusses aspects of the 1989 raid as well as the legal charges 

resulting from the raid (Swenson 1995).

• A detailed published response from a manager in the Environmental Department who was 

apparently an individual of investigative interest during the raid (An Insider’s View of Rocky 

Flats: Urban Myths Debunked) (Hobbs and Warden 2010).

• A petitioner representative provided the DOE’s Initial Agency Decision for Case 

No. VWA-0031 dated August 6, 1999 (DOE 1999a). This case involves a complaint from a 

former RFP worker alleging management reprisals after disclosures of possible health and 

safety violations and site mismanagement. The scope of the allegations is outside the period 

and location under evaluation in this attachment and does not affect NIOSH dose 

reconstructions for RFP.

Although it initially appeared that many of the site documents relevant to the period of the FBI’s raid 

were sealed in files associated with the litigation, coordination with individuals who manage RFP site 

records resulted in access to additional pertinent documents that support this assessment. This 

attempt included efforts to locate the three remaining special assessment Tiger Team reports (for 

Management and Operations, Safety, and Legal Matters) that relate to the available DOE 

Environmental Special Assessment Team report (DOE 1989a). On August 15, 2013, the RFP records 

managers reported via e-mail that they performed an extensive search and were unable to locate the 

other three Tiger Team reports. The documents the RFP records managers did provide include:

• The complete Grand Jury Report, dated January 24, 1994, on the allegations and evidence 

from the FBI’s RFP raid (DOJ 1992). This report provides specific explanations of the RCRA 

and environmental violations associated with the raid. No personnel monitoring violations or 

other occupational radiological monitoring program deficiencies are identified in the report.
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• Copies of RFP Occupational Radiological Control Program procedures, including pre- and 

post-raid versions for determining if the site revised its procedures as a result of the raid. In 

addition, a Building 771 Laboratory Sample Handling procedure includes a series of procedure 

revisions spanning the pre- and post-raid period. The revisions support the notion that no 

procedural changes resulted from the raid (Fraser 1996).

• Section I.7 lists SRDB Ref IDs that contain examples of post-raid air sampling, bioassay 

monitoring, contamination monitoring, nasal swab, and instrument operations procedures and 

manuals. According to one interviewee, the site did not commence archiving previous 

procedures until the late 1980s; before that, obsolete versions were destroyed when new 

versions were put into place. This previous policy might explain why the above archived 

procedures only go back to the early 1990s (ORAUT 2013j).

A collection of Denver EMCBC classified and unclassified documents was obtained that would 

support the assessment of impacts during the time of the FBI raid. These documents are stored at a 

classified document storage location. The information from the captured documents supports the 

notion that any radiological program document and/or procedure revisions that occurred around the 

time of the FBI raid were made as a result of a review and assessment of the Radiological Control 

Program directed by the site managers before the FBI raid. This site’s review and assessment 

included responses to a GAO audit and Technical Safety Appraisal that appears to have included 

input from the Institute of Nuclear Power Operations (RFP 1984–1990; DOJ 1989a; FBI 1989a; DOJ 

1989b; Sanchini 1988a; 1988b; Rockwell 1989d; FBI 1989b; Sanchini 1987; 1988c, 1988d; 1988e; 

FBI 2005; Sanchini 1988f; DOE ca. 1988; DOE 1989b; Norton 1989; DOJ 1989c; FBI 1989c; FBI 

2001; Rigsby 1997; Simonson 1989; DOJ 1990; Middleton 2005; FBI 1989d).

Based on the review of the documents collected in total as part of the followup document collection, a 

conclusion was made that the reviewed information resulted in no impact that affects the ability to 

adequately reconstruct individual doses under EEOICPA. The specific information from the collected 

documents that support this conclusion includes:

• As stated by the U.S. Attorney in the RFP sentencing memorandum, there were no identified 

situations that posed an imminent threat to RFP workers, the public, or the surrounding 

environment (Swenson 1995, p. 13).

• Although the initial FBI investigation identified potential issues at RFP, the resulting FBI raid 

did not result in the same findings that initially seemed apparent to the EPA and FBI (based on 

the previous allegations and investigative characterizations that led to the raid) (Hobbs and 

Warden 2010).

• The end result was a settlement that included an agreement between parties eliminating 

further pursuit of individual indictments (Swenson 1995, pp. 12–14).

• The charges against Rockwell at RFP were specific to environmental RCRA and Clean Water 

Act Laws and the impact to the environment; the charges did not specifically call out a data 

falsification, data validity issues, or a data quality violation (Swenson 1995, p. 13–14).

I.4.1 Additional Colorado Visit for Data Capture and Interviews

Additional contacts that might have pertinent or contain information relevant to the issues being 

investigated as part of this followup were identified from ABRWH Working Group meetings. One 
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individual (ORAUT 2013l) was scheduled for an unclassified telephone interview (followed by a 

classified telephone interview and followup call). A second individual (ORAUT 2015a) specifically 

requested a classified interview and was interviewed in a secure setting at a location in Idaho Falls. 

The data and information from these interviews are discussed here.

• The individual involved in the unclassified interview provided information about involvement in 

shredding documents (as pertinent to this petition and the EEOICPA dose reconstruction 

process). The individual indicated that the direction for document destruction came from RFP 

management. The individual discussed feeling uncomfortable about destroying the 

documents because they were originals containing information on RFP operations, including 

monitoring data and incident reports. Several potential interviewee names were provided 

during the interview and the individual provided examples of the documents that were 

destroyed as part of the process (examples were received after the interview). The individual 

also stated a desire to participate in a secure interview to discuss other information relating to 

the document destruction process that might be classified.

A followup interview was performed in February 2014 (ORAUT 2014f) to collect additional 

information and documentation that was discussed during the initial interview with the first 

interviewee. A secure telephone interview was scheduled and completed in May 2014 at the 

request of the interviewee (ORAUT 2014g) and all information discussed during that interview 

was cleared for uncontrolled use and dissemination. This secure telephone interview resulted 

in information from the initial interview being restated with additional information about what 

was believed to be the projects related to the documents that were destroyed. A final followup 

telephone discussion occurred in June 2014 (ORAUT 2014h) to discuss the results of the 

document and/or information reviews and information provided during the preceding 

interviews.

The findings of the discussions and information reviews provided by this interviewee are as 

follows:

– While the documents could have been field surveys, it does not appear that those 

surveys have an impact on the ability to bound or reconstruct dose for the class as long 

as the personnel monitoring data exist. Based on a review of some of the files that 

were provided as examples of documents that the interviewee believed were 

destroyed, the records were found to exist in the associated personnel files in NOCTS. 

Therefore, it appears unlikely any dose records were destroyed.

– A potential additional interviewee was mentioned during this interview who was already 

on the list of desired interviewees; however, there was no success coordinating with 

this individual to schedule an interview.

• The individual interviewed during the classified interview at Idaho Falls (as part of the 

December 2013 Denver data capture trip) provided information about the assessment of data 

falsification at RFP. This individual did confirm the separation of the Environmental and 

Occupational Radiological Analysis programs within Building 123. The individual relayed 

information pertinent to the ability to reconstruct radiation dose about bioassay and personnel 

monitoring and specifically involved information about penciling-in and changing dosimeter 

readings and misplaced or lost bioassay samples, as well as contamination incidents and 

safety issues at RFP.
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Some specific information from this interview is:

– Penciling-in dosimeter information. The interviewee discussed a situation where 

dosimetry technicians wrote dose rate information in reports in pencil, which would 

allow RFP management to later direct changes to keep the production going at RFP. It 

was relayed that this issue was addressed through a grievance at RFP.

Based on NIOSH professional judgment and field experience, the issue of penciling-in 

information appears to refer to radiological field survey records that directly relate to 

ongoing production operations. The only dosimetry information that might be included 

in such field surveys would be from direct-reading dosimeters or personal ion 

chambers (PICs, a.k.a. pencil dosimeters). While field survey information is used for 

comparison purposes in the performance of EEOICPA dose reconstruction, the primary 

and most important source of radiological information for the purpose of individual dose 

reconstruction is the individual TLD dosimetry and bioassay information. TLD and 

bioassay analyses are performed in a laboratory and not documented in the field, in 

contrast to the surveys and reports discussed in the claim raised by this individual. 

Therefore, it is not expected that the original handwritten documents that the 

interviewee referred to in the destruction process are about an individual’s TLD or 

bioassay results (with electronic readouts; Inkret undated; RMRS 1999; RFP 1994).

– RFP bioassay issues. The interviewee relayed personal concerns with the bioassay 

program and relayed a specific concern about bioassay sample analysis results (false 

positives and [statistical] variations in bioassay results (see the similar second 

interviewee claim and response in Section I.4.2 below. While no other specific 

concerns about falsification of records were brought up, the interviewee did raise other 

general concerns about bioassay sample handling and processing.

This concern does not raise issues that invalidate the use of personnel bioassay data 

in the performance of dose reconstructions under EEOICPA. The dose reconstruction 

process accounts for the potential for missed doses and incorporates methods that are 

favorable to the claimant. Therefore, the conclusion was made that this issue does not 

affect the dose reconstruction process.

– Personnel contamination problems and other contamination incidents. The interviewee 

brought up a significant number of radiological and some nonradiological 

contamination incidents and exposure-control issues in reference to the overall safety 

program at RFP.

While contamination incident and survey data are used to supplement the personnel 

monitoring data in the performance of dose reconstructions under EEOICPA, 

personnel monitoring data are considered the primary data sources for the process. 

Therefore, the conclusion was made that this issue does not affect the dose 

reconstruction process. 

– Tritium bubblers, neptunium, MgTh alloy, and CML. The interview provided some 

information on these other issues at RFP.

All of these issues are addressed in various sections of this TBD.
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I.4.2 Site Visit and Data Capture Interviews after Denver Visits

Five additional potential interviewees with information on the subject were identified from ABRWH 

Working Group meetings, petitioner information, and individual interviews. Attempts were made to 

schedule interviews with each of them. Two of the five were contacted and interviewed, one about 

data falsification and one about the FBI raid; the remaining three either did not agree to be 

interviewed or did not return messages requesting an interview. Only the data and information from 

interviews as part of the effort are discussed in this Attachment.

I.4.2.1 Data Falsification Interviewee

The individual interviewed about data falsification (ORAUT 2014i) had served as a [redacted] and 

[redacted] during employment at RFP. The interviewee confirmed that Environmental Radiological 

Program changes did occur as a result of the FBI raid. The interviewee also brought up concerns 

associated with personal radiological monitoring records and the radiological monitoring program at 

RFP, which documented varying positive and negative bioassay results in an individual’s dose 

records.

No information from this interview supported the allegation of document destruction activities at RFP. 

While the individual discussed concerns with the implementation of radiological limits and controls as 

well as dose reporting during employment, up to the point of the implementation of the DOE 

Radiological Control Manual at the site (late 1980s to early 1990s) and the FBI raid, there were no 

identified effects on the ability to bound dose for the portion of the class of RFP workers being 

assessed for this attachment. It was confirmed that the focus of the FBI raid was environmental 

radiological issues and that program changes did occur as a result of the raid. The other concerns 

that were relayed were associated with personal radiological monitoring records and the 

documentation of the statistical or sample-counting variations that can produce positive and negative 

bioassay results associated with the analysis of a potential exposure situation in an individual’s dose 

records. This interviewee specifically mentioned situations in which the site would increase limits to 

preclude exceeding them. In other cases, the interviewee believed that while reported results would 

indicate “no data available” but that that there were results that were available (i.e., based on their 

knowledge of the area survey results, personnel should have a recorded dose value). The 

interviewee indicated that when there were problems with the dosimetry readings, the individuals 

would receive an average of his coworkers’ doses. The interviewee believed that there are cases 

where these averages are incorrect. These issues were assessed by the ORAU Team, which 

considered the interviewee information, reviewed worker files for comparable or corroborating 

situations, and assessed the effects of the verified issues on the completion of the EEOICPA dose 

reconstruction process. The assessment specifically focused on potential negative effects on 

individual dose reconstructions based on the interviewee’s concerns. The ORAU Team did not 

identify situations or issues that would affect the ability to reconstruct dose for the RFP worker class 

being assessed as part of this attachment.

I.4.2.2 FBI Raid Interview and Followup Information Reviews

The SEC-00192 RFP petitioner identified the lead FBI agent involved in the 1989 FBI raid and 

provided the applicable contact information for the former agent (ORAUT 2014j). The agent was 

interviewed and provided a significant number of papers and documents about the FBI’s raid of RFP 

in 1989. Many of the documents were the property of other government agencies or entities and 

required release by the applicable agency general counsel before they could be used or referenced in 
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a EEOICPA project document. The following subjects and issues were identified during the interview 

and subsequent followup with the agent:

• The contention of the agent was that there was an additional August 1989 aerial MSS at RFP 

in addition to the one in June and July of 1989. The agent submitted a FOIA to DOE 

requesting that additional MSS information be provided. DOE completed searches and 

discovered no additional survey information. The agent submitted an appeal of that response, 

which was subsequently denied by the DOE Office of Hearings and Appeals citing that the 

search was adequate and no additional information is available from Legacy Management 

(DOE 2012).

• Part of the response to the FBI raid at RFP was for DOE to initiate its own special assessment 

of the site using Tiger Teams, which paralleled the operations of the FBI raid. The report that 

is currently available is titled Assessment of Environmental Conditions at the RFP, written by 

the U.S. DOE Special Assignment Environmental Team in August 1989 (DOE 1989a). In 

addition, three other reports for management and operations, safety, and legal matters are 

identified; however, those reports are not available and do not appear to exist based on 

extensive coordination and searching with the FBI agent as well as DOE and Legacy 

Management in Denver.

• There is a contention that the flyover data indicate the presence of the isotopes 137Cs and 90Sr, 

which is used to imply that an unreported criticality occurred at RFP. This relates to the 

contention from the agent about the existence of an August 1989 flyover MSS.

No August 1989 flyover survey has been found. In addition, no specific information was found 

that supported a criticality event. Individuals were interviewed to assess the criticality claim 

with the following results:

– Subsequent interview discussions and a report obtained from an interviewee do not 

corroborate the occurrence of a criticality at RFP. This includes the “1989 Criticality 

Safety Assessment at Rocky Flats” of the Assessment of Environmental Conditions at 

the RFP (DOE 1989a, p. 277). The Assessment Team found no indication that a 

criticality accident ever occurred at RFP.

– An interviewee provided a report after his interview that concludes there is no evidence 

to support that a criticality occurred at RFP (Scientech 1989). The report does go on to 

identify issues with the Criticality Safety System and Program at the site and includes 

fixes for the issues, but no conclusive evidence, monitoring or other, was identified to 

support the occurrence of a criticality incident.

• The agent discussed the Waste Stream Identification and Characterization (WISC) Reports 

and identified them as a major source of information to support the raid at RFP. The entire set 

of WISC reports was obtained and reviewed for applicability to the SEC evaluation and dose 

reconstruction processes for RFP. While the information might have been useful to 

characterize conditions and issues at the time of the raid, no information to support evaluation 

of or reconstruction of dose at RFP were identified (Rockwell 1987i to 1987ah).

• As part of the response to the interview, the agent provided over a thousand pages of 

documents. Other than the documents specifically called out in this section, the remainder are 

considered the property of another government entity or legal interest, and coordination 
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between NIOSH, Centers for Disease Control and Prevention, Office of General Counsel, and 

the legal representatives of the other organizations is required for approval to obtain and 

reference the applicable documents. The documents applicable to the assessment presented 

here include: (1) RFP personnel interviews (by both the FBI and EPA) at the time of the raid; 

(2) FBI research leading up to the raid; (3) Building 123 information associated with the FBI 

investigation and warrants; and (4) 1988 aerial survey documentation of the RFP site. The 

documents that are the property of the FBI were released by the FBI General Counsel on 

January 26, 2015; the general counsel release approval for the DOE and EPA documents was 

obtained on April 21, 2015:

– FBI Interviews. FBI (1991a to 1991c, 1989e to 1989i).

– EPA Interviews. RFP (1975–2004, pp. 1214, 1215–1222, 1223–1224, 1225–1226, and 

1227).

– FBI research documents leading up to the raid. RFP (1975–2004).

– Building 123 notes about FBI investigation and warrants. DOJ (1989a to 1989c).

– 1988 aerial survey of the RFP site. Author unknown (1988); note that no 1989 survey 

identifying radiological levels to support a criticality has been found.

• The agent discussed and provided some notes from a personal notebooks of an RFP manager 

about the RFP radiological program (Sanchini 1988b). Additional portions of the notebooks 

were obtained in subsequent data capture efforts (Sanchini 1987, 1988a, 1988c to 1988e). 

While some notes discussed needed improvements in the radiological program and identified 

program confidence issues, they do not indicate program deficiencies that represent an 

inability to bound dose.

• The agent’s response to the documented interview included the identification of a significant 

number of names and contacts associated with the raid. Of these names, five individuals’ 

contact information was identified and three of the five were successfully contacted and 

agreed to an interview. The three interviews are summarized below:

– The first interview (ORAUT 2014k) was with the DOE Senior Site Representative at the 

time of the FBI raid. This individual indicated awareness that the issues were 

associated with environmental violations and not with occupational radiological 

monitoring violations. The individual also discussed issues that involved personnel 

complaints about availability of monitoring records, which the interviewee deemed 

hearsay.

– The second interview (ORAUT 2014l) was with a Rockwell and DOE Office 

spokesperson. During the raid, the individual mainly dealt with the news media and 

provided responses about the raid. Other than that, the individual had no direct 

involvement in dealing with the raid. Part of the issue that resulted in the raid involved 

late night burning and operations of a waste incinerator. All the issues of which the 

interviewee was aware were environmentally related and had no occupational 

radiological components. This individual did have some personal records at home and 

coordination is occurring to obtain copies of any pertinent information.
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– The third interview (ORAUT 2014m) was with an individual who had never been an 

employee or contractor at RFP. The only two associations this person had with RFP 

was as part of a Governor’s panel to investigate unreported criticalities and serving as 

part of a NIOSH Health Surveillance Program. As part of the response, the individual 

reported that no indications were identified that supported the occurrence of a criticality 

at RFP. The individual provided a copy of the report that followed. As part of the 

investigation, the individual looked for anomalies in personnel doses as well as flyover 

data. The investigation revealed no suspicious information and indicated that there 

was good continuity in the data. The Health Surveillance Program also included the 

assessment and revision of site personnel neutron doses.

The information from the FBI agent, including interview information (from the agent and other 

knowledgeable individuals identified by the agent) and associated documentation, support the 

idea that the basis of the FBI raid was RFP environmental issues. While some information 

collected and assessed at the time of the raid does cross over into occupational radiological 

issues (including the RFP manager personal notebooks), nothing was discovered that 

supports a data falsification or destruction issue that would affect the ability to reconstruct 

dose for the RFP worker class being assessed as part of this attachment. The review of the 

flyover and MSS information indicates that there might have been support for the FBI raid from 

an environmental perspective. However, these documents and files provided no evidence or 

information that disputes the ability to bound RFP worker dose under the EEOICPA program. 

The claim of an unreported criticality incident at RFP was investigated from several points of 

view in the documents that were reviewed, as well as during the interviews of knowledgeable 

individuals; no supportive evidence of a criticality incident was found. The conclusion was 

made that no information about this issue affects the ability to reconstruct individual dose 

under EEOICPA.

I.4.3 Review of Petitioner-Identified Technical Safety Appraisal Issues

As part of the followup issues raised by the petitioner during the Idaho Falls ABRWH meeting, a GAO 

report was referenced (GAO 1988) that, in turn, referenced an RFP Technical Safety Appraisal (TSA) 

(DOE 1988b). The petitioner specifically cited the following statements from the TSA:

• DOE 1988b, pp. 10–11. For example, radiation monitoring is adversely affected by poor 

quality instrumentation, inadequate calibration techniques, and improper use of equipment. 

The Radiological Health Quality Assurance Program is ineffective as evidenced by some of 

the preceding concerns, failure to comply with DOE-prescribed standards, and deficiencies in 

maintaining exposure records and tracking bioassay samples.

• DOE 1988b, p. 174. During the past few weeks, several SAAM’s were turned off without 

notifying either radiation monitoring or the instrument technicians. These instruments were 

operational when turned back on. There is no electronic method to automatically display their 

operational status in the monitoring office.

• DOE 1988b, p. 180. The health physics instruments used for personnel protection do not all 

conform to appropriate performance requirements of applicable standards.

The RFP TSA was reviewed for information applicable to the assessment of personnel exposure 

information, records, and exposure tracking. The appraisal contains information about the radiological 

interface between the RFP radiological protection personnel and other plant organizations as 
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described in the report; Section L focuses specifically on performance objectives, findings, and 

concerns about radiological protection (DOE 1988b, pp. 161–191). The concerns identified in 

Section L are later listed in Appendix C, “List of Concerns,” under “L. Radiological Protection” (DOE 

1988b, pp. 243–245).

The reviewed information was assessed to determine if the TSA findings have any substantial effects 

on the dose reconstruction process. In the “Major Problem Areas” summary of the appraisal (see the 

first bullet above; DOE 1988b, pp. 10–11), there are two noted deficiencies that directly relate to the 

primary dose reconstruction data sources: (1) proper use of radiation monitoring equipment 

[dosimetry], and (2) maintaining exposure records and tracking bioassay samples.

• The first issue arises from the observation of inadequate external dosimeter placement on a 

person’s body. The provided example was truck drivers wearing dosimeters on their chest 

when radioactive materials were behind them (Concern RP.5-2; DOE 1988b, pp. 171, 244). 

Although any determinations on the reconstruction of dose for offsite shipments and/or 

couriers is outside the scope of this review, the assessment of external exposure geometry 

issues related to this topic was performed as part of previous evaluations, specifically in the 

SEC-00030 RFP evaluation report (NIOSH 2006). As such, the external dose adjustment to 

account for exposure geometry is considered an individual dose reconstruction issue to be 

handled on a case-by-case basis.

• The second issue involves internal dosimetry and bioassay and is associated with tracking 

routine and special bioassays, the need for an independent audit program, and the need for a 

program to compare in vivo and in vitro results for particular employees. These issues do not 

speak to a lack of internal dosimetry data, but rather methods to better track and compare 

them.

The remaining issues in the statements cited by the petitioner in the bulleted paragraphs above are 

about field monitoring surveys and instrumentation associated with RFP operations; they are not 

specific to the personnel internal or external radiological monitoring data used on the NIOSH dose 

reconstruction project.

Based on a review of the RFP TSA in its entirety, there was no identified effect on radiological 

personnel monitoring data that are used to support bounding doses or dose reconstruction for the 

RFP worker class assessed.

I.4.4 Review of Petitioner-Identified Data Falsification Issues Report

The petitioner-provided document (DOE 1999b) in support of data falsification issues is associated 

with a complaint filed by an RFP worker under the DOE Contractor Employee Protection Program, 

10 CFR Part 708. In the complaint, the worker contended that reprisals by the employer were taken 

after certain disclosures were made concerning possible health and safety violations and 

mismanagement at the RFP site. It was found that the employer did take acts of reprisal against the 

employee prohibited under 10 CFR § 708.5, and that the employee was entitled to remedial action 

from the employer.

The petitioner-provided report centers on an allegation of records falsification involving mislabeling 

waste for shipment off site. The findings discuss allegations of potential retaliation for protected 

activities; however, it only contains potential effects in terms of environmental dose. No information 
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affects the ability to reconstruct radiation dose with sufficient accuracy under the EEOICPA Program 

for the RFP worker class assessed.

I.4.5 Additional Information from RFP Manager Personal Notebooks

On January 28, 2015, the petitioner provided excerpts from the notebook of the previously discussed 

RFP manager who made notes on the radiological program and who was present at the time of the 

FBI raid. The petitioner made reference to the notebook statement that the program data was “bad,” 

but the statement in the notebook is in reference to collecting samples but not counting them 

(Sanchini 1998e). On March 30, 2015, the petitioner provided additional excerpts from the same 

notebooks denoting that the notebooks indicate “questionable” operations in the Building 123 

bioassay laboratory (Sanchini 1986–1988, 1988–1989).

While some information collected and assessed at the time of the raid does cross over into 

occupational radiological issues (including the RFP manager personal notebooks), nothing was 

discovered that supports a data falsification issue that would affect the ability to reconstruct dose for 

the RFP worker class assessed.

I.5 ASSESSMENT OF AVAILABLE PERSONNEL RADIOLOGICAL MONITORING DATA

As part of the followup to the assessment for this attachment, a review was performed of the quantity 

of available personnel radiological monitoring data available. A conclusion was made that there exists 

a sufficient quantity of individual external monitoring data to support the assessment of RFP 

personnel external doses. 

Specifically, there are sufficient individual external dosimetry TLD data to support reconstruction of 

RFP personnel external dose for after 1983 using the methods defined in the main text. Figure I-1 

shows the number of workers who were badged over various years. The solid line derives from 

EEOICPA claim data; the broken line shows the results of a similar analysis on the RFP electronic 

files for the entire plant population. The difference is due to discrepancies in termination dates in the 

electronic files. These dates are verified as worker files are reviewed, providing a better 

understanding of the badging process as well as a more accurate individual worker record.

For RFP personnel internal doses, an assessment of the quantity of data available from after 1983 

was performed. The assessment involved a review of the internal RFP co-exposure study of data 

from 1983 to 1988 (which was later extended beyond 1988; see Attachment D). The review identified 

the results shown in Table I-1.

Information on the variances in numbers of bioassay samples notes that there was a significant 

increase in employment in 1984 (to a peak of 5,990 in 1984), and that plutonium processing ceased in 

1990. While no clear correlation can be made for 1985 to 1986 and 1988 to 1989, the 1985 date does 

correspond to the incorporation of new air sample filter media in the air monitoring program, and the 

1989 date corresponds to EG&G taking over the site and the occurrence of the FBI/EPA raid at RFP.

Based on this information and the information in the preceding sections of this assessment, there is 

no issue associated with the FBI raid or issues about data falsification or invalidation that affect the 

data from after 1983 in a way that would preclude individual dose reconstructions with sufficient 

accuracy under EEOICPA.
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Figure I-1. Percent of workers badged for dosimetry (NIOSH 2006, Figure 6-1). 

Table I-1. Available internal sample data, 1980 to 1995. 

Year

Total

urine samples Pu urine U urine Pu fecal

Total

in vitro samples

Total

lung counts

1980 5,906 5,628 278 0 5,906 4,050

1981 6,349 5,991 358 0 6,349 4,384

1982 8,527 7,109 1,418 0 8,527 4,876

1983 9,181 7,506 1,675 0 9,181 6,052

1984 10,468 8,830 1,638 0 10,468 6,777

1985 10,381 8,767 1,614 0 10,381 5,922

1986 7,075 5,893 1,182 0 7,075 5,460

1987 3,979 3,429 550 0 3,979 4,966

1988 4,808 4,415 393 0 4,808 5,735

1989 6,977 4,777 2,200 78 7,055 3,618

1990 5,172 4,168 1,004 69 5,241 1,799

1991 495 483 12 225 720 638

1992 960 762 198 758 1,718 553

1993 1,550 1,197 353 368 1,918 663

1994 3,325 2,722 603 770 4,095 892

1995 2,106 1,688 418 69 2,175 239

I.6 GENERAL SUMMARY AND CONCLUSIONS

This section provides a general summary of the sequence of topics addressed above:

• A review of the interviewee allegation relevant to data falsification and data invalidation in 

Building 123 was performed. The interviewee made statements about the inadequacy of fume 

hoods and the improper handling and preparation of environmental, bioassay, fecal coliform, 

and stack samples. From a radiological perspective, no scientific basis was found for 

concluding that the issues raised about environmental samples would compromise radiological 

count results, and the reviewed information does not corroborate a link between the 

environmental and occupational radiological programs. Nevertheless, four individuals were 
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interviewed with potential related knowledge or information who confirmed that the focus of the 

assessment and the FBI raid was very specific to environmental impacts and monitoring. 

There is no relationship between the identified environmental findings or deficiencies to the 

worker monitoring program, and no such formal allegations were made.

• A review of An Insider’s View of Rocky Flats: Urban Myths Debunked was performed (Hobbs 

and Warden 2010). The authors implied that the FBI raid found no issues with worker 

protection or the worker monitoring program; the only violation cited for RFP was an 

environmental release. Based on the DOE Special Assessment report, the focus of the 

assessment was very specific to environmental impacts and monitoring (DOE 1989a). 

Although the report discusses data QA and validation issues with the analytical laboratories in 

Building 881 and Building 123, the conclusion was made that no situations were identified in 

which falsification or invalidation of data would affect the ability to perform dose reconstruction 

under EEOICPA.

• RFP site records managers provided access to additional pertinent documents that support 

this assessment, including the complete Grand Jury Report, RFP Occupational Radiological 

Control Program procedures, and examples of post-raid air sampling, bioassay monitoring, 

contamination monitoring, nasal swab, and instrument operations procedures and manuals. 

Based on its review, the ORAU Team concluded that the information does not affect the ability 

to adequately reconstruct individual doses under EEOICPA. There were no identified 

situations that posed an imminent threat to RFP workers, the public, or the surrounding 

environment. The FBI raid did not result in the same findings that initially led to the raid. 

Although Rockwell pled guilty and paid a fine, it appears the settlement was based on the 

company’s desire to close the prolonged litigation. Furthermore, the charges against Rockwell 

were specific to environmental RCRA and Clean Water Act Laws and environmental impacts; 

the charges did not specifically call out a data falsification, data validity issues, or a data 

quality violation.

• One individual provided information about involvement in shredding documents. The ORAU 

Team concluded that while the documents being destroyed could have been some kind of field 

surveys, it does not appear that those surveys have an effect on the ability to bound or 

reconstruct dose, as long as the personnel monitoring data exist. Based on a review of some 

of the files that were provided as examples of documents that the interviewee believed were 

destroyed, it was determined that the records did exist in the associated personnel files in 

NOCTS; therefore, those files were not destroyed.

• An interviewee raised the issue of dosimetry technicians writing down dose rate information in 

pencil, which would allow RFP management to later direct changes to keep the production 

going. Based on professional judgment and field experience, the issue of penciling-in 

information appears to refer to radiological field survey records that directly relate to ongoing 

production operations. The only dosimetry information that might be included in such field 

surveys would be from direct-reading dosimeters or personal ion chambers, which are used for 

comparison purposes in EEOICPA dose reconstruction. The primary and most important 

source of radiological information for individual dose reconstruction is the individual TLD 

dosimetry and bioassay information. TLD and bioassay analyses are performed in a 

laboratory and not documented in the field.

In addition, this interviewee relayed concerns about bioassay sample analysis results (false 

positives and [statistical] variations); bioassay sample handling and processing; personnel 
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contamination and contamination incidents; and issues about tritium bubblers, neptunium, 

MgTh alloy, and the CML. The ORAU Team concluded that the concerns about bioassay do 

not raise any issues that invalidate the use of personnel bioassay data in the performance of 

dose reconstructions under EEOICPA. The dose reconstruction process accounts for the 

potential for missed doses and incorporates methods that are favorable to the claimant. 

Contamination incident and survey data are used to supplement the personnel monitoring data 

in the performance of dose reconstructions under EEOICPA, but personnel monitoring data 

are considered the primary data sources for the process. All of the issues concerning tritium 

bubblers, neptunium, MgTh alloy, and the CML are outside the scope of this attachment and 

are addressed in other topic-specific documents.

• Two of five potential interviewees were interviewed based on information identified from 

ABRWH Working Group meetings, petitioner information, and individual interviews. The 

remaining three either did not agree to be interviewed or did not return messages requesting 

an interview. 

– The first interviewee confirmed that Environmental Radiological Program changes did 

occur as a result of the FBI raid and brought up concerns associated with personal 

radiological monitoring records and the RFP radiological monitoring program, which 

documented varying positive and negative bioassay results in an individual’s dose 

records. The ORAU Team concluded that no information provided during this interview 

supported the allegation of document destruction activities at RFP or affected the 

ability to bound dose for the portion of the class of RFP workers being assessed.

– The second interviewee was the lead FBI agent in the 1989 FBI raid, who provided a 

significant number of papers and documents about the raid. Many of these documents 

are awaiting release by the applicable agencies. The agent contended that there was 

an additional August 1989 aerial MSS at RFP in addition to the one in June and July of 

1989, and that the flyover data indicate the presence of the isotopes 137Cs and 90Sr, 

which was used to imply that an unreported criticality occurred. NIOSH could locate 

neither the August 1989 flyover survey nor evidence supporting a criticality event. 

Other individuals were interviewed to assess the criticality claim and could produce no 

corroborating evidence. The information from the FBI agent supports the idea that the 

basis of the FBI raid was environmental issues. The ORAU Team concluded that the 

interview, documents, and files provide no evidence or information that disputes the 

ability to bound RFP worker dose under the EEOICPA program.

• A review of the entire RFP Technical Safety Appraisal (TSA; DOE 1988b) was made after the 

petitioner cited three statements concerning poor-quality instrumentation, standards, and 

record tracking. In the “Major Problem Areas” summary of the appraisal, there are two noted 

deficiencies that directly relate to the primary dose reconstruction data sources for NIOSH 

under EEOICPA: (1) proper use of radiation monitoring equipment [dosimetry]; and 

(2) maintaining exposure records and tracking bioassay samples. 

– The ORAU Team concluded that any determinations on the reconstruction of dose for 

offsite shipments or couriers is outside the scope of this review. However, the 

assessment of external exposure geometry issues related to this topic was performed 

as part of previous evaluations and is incorporated into other sections of this TBD. As 

such, the external dose adjustment to account for exposure geometry is considered an 

individual dose reconstruction issue to be handled on a case-by-case basis.
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– The second issue involves internal dosimetry and is associated with tracking routine 

and special bioassays, the need for an independent audit program, and the need for a 

program to compare in vivo and in vitro results for particular employees. The ORAU 

Team concluded that these issues do not speak to a lack of internal dosimetry data but 

rather to methods to better track and compare these personnel monitoring data. Based 

on its review of the RFP TSA in its entirety, there is no identified effect on radiological 

personnel monitoring data that are used to support bounding and/or reconstructed 

dose for the RFP worker class assessed.

• A petitioner-provided report was reviewed centering on an allegation of record falsification 

involving mislabeling waste for shipment off site. The findings discuss allegations of potential 

retaliation for protected activities; however, it only contains potential impacts in terms of 

environmental dose. The ORAU Team concluded that none of this information affects the 

ability to reconstruct radiation dose with sufficient accuracy under the EEOICPA Program for 

the RFP worker class assessed.

• The petitioner provided excerpts from the notebooks of an RFP manager who made notes on 

the radiological program and who was present at the time of the FBI raid. The ORAU Team 

concluded that while some information collected and assessed at the time of the raid does 

cross over into occupational radiological issues, nothing was discovered that supports a data 

falsification issue that would affect the ability to reconstruct dose for the RFP worker class 

assessed. 

• A review was performed of the quantity of personnel radiological monitoring data available at 

the time of this assessment. The ORAU Team concluded that there exists a sufficient quantity 

of individual external monitoring data to support the assessment of RFP personnel external 

doses. 

Based on this information and the information in the preceding sections of this attachment, issues 

associated with the FBI raid or data falsification or invalidation do not affect the post-1983 data and do 

not preclude individual dose reconstructions under EEOICPA.

I.7 EXAMPLES OF POST-RAID PROCEDURES AND MANUALS

The following SRDB Ref IDs contain examples of postraid air sampling, bioassay monitoring, 

contamination monitoring, nasal swab, and instrument operations procedures and manuals that were 

reviewed for this analysis: 126927, 126926, 126925, 126924, 126923, 126922, 126920, 126919, 

126917, 126916, 126913, 126906, 126902, 126901, 126899, 126896, 126895, 126892, 126890, 

126888, 126887, 126886, 126885, 126884, 126883, 126882, 126881, 126880, 126878, 126877, 

126876, 126874, 126873, 126872, 126871, 126869, 126868, 126867, 126866, 126864, 126863, 

126836, 126833, and 126831.