Page 1 of 180

DOE Review Release 08/26/2020

Hanford Site – Occupational Internal Dose

ORAUT-TKBS-0006-5 Rev. 07

Effective Date: 09/01/2020

Supersedes: Revision 06

Subject Expert(s): Fred L. Duncan and Matthew G. Arno

Document Owner 

Approval: Signature on File

Fred L. Duncan, Document Owner

Approval Date: 06/19/2020

Concurrence: Wade C. Morris Signature on File for

John M. Byrne, Objective 1 Manager

Concurrence Date: 06/19/2020

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 06/18/2020

Concurrence: Signature on File

Matthew C. McFee, Deputy Project Director

Concurrence Date: 06/22/2020

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-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 2 of 180

 

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

10/15/2003 00 New document to establish TBD for occupational internal dose –

Section 5. First approved issue. Initiated by Edward D. Scalsky.

11/24/2004 01 Approved revision initiated to discuss historical limits and tolerance 

dose; discussion of separations plant (1944-46) and 231-Z (1945-

46); clarified intakes in the 300 Area uranium fabrication and the 

laundry facilities; expanded information on 241Am and 241Am MDAs 

for 1946 and 1967-6/1969; added Section 5.2.4.1 on assignment of 

tritium doses; added Table 5.2.5-5 (MDAs for nonroutine uranium 

excreta analyses); revised Table 5.2.6-1 (routine fission product 

urinalysis detection levels); added discussion of 214Bi and 208Tl in 

whole-body counting (Section 5.3.1); expanded Section 5.7 on 

unmonitored workers to provide additional discussion, tables, and 

instructions to dose reconstructors for specific areas on the site; 

included listing of references; plus corrected miscellaneous typos, 

and table number changes. Incorporates formal internal and NIOSH 

review comments. Initiated by Edward D. Scalsky.

06/22/2007 02 Approved revision initiated to revise or add information in Section 5.

Section 5.1: updated tolerance air concentrations and assigned 

intakes for 1944-48; added summary of Ikenberry report on thyroid 

monitoring 1945-46; added fission product intakes for 1947. Section 

5.2: added statement on types of urinalysis kits normally used.

Section 5.2.1: added information about pure 238Pu work in 325 Bldg.; 

added statement to use 5-yr-old Pu mixture for years prior to 1951.

Section 5.2.2: added statement about 241Am recovery operation in 

PFP in 1960-70s. Section 5.2.3 and 5.2.5 were changed to add 

statements about curium and uranium work, respectively. Section 

5.2.5 changed period for 4 μg/L reporting level to 1974; increased 

239Pu contaminant level in RU to 0.8 nCi/g U; added information 

about 233U production. Section 5.2.8: added information about 

promethium use. Section 5.2.10: added information about neptunium 

urinalyses. Section 5.2.11: added information about thorium 

urinalyses and 35S urinalyses. Revised or added information to 

Section 5 tables. Introduction: added new boilerplate. All table 

numbers were reformatted. Section 5.1: removed reliance on air 

samples and referenced co-worker study (OTIB-0039), Section 5.2:

added statement on types of urinalysis kits normally used. Sec. 5.2.1: 

added information about pure 238Pu work in 325 Bldg.; added 

statement to use fresh Pu for 1944-49 and 5-yr-old Pu mixture for 

1950-54. Sec. 5.2.2: added statement about 241Am recovery 

operation in PFP in 1960-70s. Sec. 5.2.3: added statement about 

244Cm recovery in 325 Bldg. Sec. 5.2.4: added information about 

tritium in the 400 Area drinking water. Sec. 5.2.5: Table 5-13: raised 

Pu-alpha contamination level to 0.8 nCi/g U; added statement about 

work with uranium; changed period for 4 μg/L reporting level to 1974; 

added information about 233U production. Sec. 5.2.8: added some

information about promethium use. Sec. 5.2.9: removed discussion 

about background levels of 210Po. Sec. 5.2.10: added some 

information about neptunium urinalyses. Sec. 5.2.11: added 

information about thorium urinalyses and 35S urinalyses.
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 3 of 180

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

06/22/2007 

(continued)

02 

(continued)

Sec. 5.3.1: revised Table 5.3.1-1 (5-22) to clarify MDAs, decision 

levels and reporting levels. Sec. 5.3.4: totally new section on thorium 

exposure and thorium monitoring. Sec. 5.4: replaced the old mixture 

rules with use of Table 5.7-2 (5-25); Sec. 5.6.2: (chronic intakes) 

deleted entire section. Sec. 5.7 (now 5.6): added information about 

tritium in the 400 area drinking water; added reference to OTIB-0039 

and revised Table 5-25 accordingly. Moved old Appendix D into this 

document as Attachment A; took out all the tables on chronic intakes.

Constitutes a total rewrite of the document. Incorporates NIOSH 

formal review comments. Many grammatical and format changes 

throughout. Table 5-12: removed the uranium specific activity data 

for DU and natural U; added statement to use IMBA defaults to 

footnote a. Section 5.3.4: added table showing buildings with thorium 

exposure; added text and table explaining adjustment factors for 

whole body measurements of 232Th and associated intakes; added 

thoron doses for workers processing thorium in 3722 and 3732 

buildings. Section 5.6: clarified possible absorption types for Pu and 

U in the specific-facilities section; added thoron to Table 5-30; 

revised tritium intakes for unmonitored workers due a coworker 

analysis of tritium doses. New Tables 5-27 and 5-28 providing 232Th 

intakes based on whole body counting at long times after end of 

chronic intakes. Section 5.3.4: revised thoron WLMs, removed 

reference to thoron dose with a reference to TIB-011, clarified use of 

OTIB-0039 for unmonitored thorium intakes. Section 5.6: removed 

references to 6-mrem tritium dose and stayed with intake values; 

clarified how the tritium intake values were determined for the 1955-

60, 1970-71, and 1973-83 time periods; corrected a math error 

resulting in slightly increased intakes for 1984, 85, and 86. Includes 

Attributions and Annotations and additional references. Incorporates 

formal internal review comments for Attributions and Annotations 

section. This revision results in an increase in assigned dose and a 

PER is required. Training required: As determined by the Task 

Manager. Initiated by Donald E. Bihl and Edward D. Scalsky.
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 4 of 180

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

01/07/2010 03 Revision to add new SEC statement in Section 5.1. All references to 

OTIB-039 were removed. OTIB-039 was added as Attachment C. All 

mention of assigning intakes for unmonitored workers for Np, Po210, or thorium were removed. Attachment A was split with REX 

codes becoming a new Attachment B and tolerance limits remaining 

as Attachment A. Table 5-1 on bioassay codes were moved to 

Attachment B. Text on records missing in REX was added to Section 

5.2. In Section 5.2.1, text was added on change in bioassay 

frequency circa 1960. Text on history of Am exposure and start of 

bioassay was added in Section 5.2.2. Text on history of Cm exposure 

was added to Section 5.2.3. Text on history of tritium exposure and 

discussion of metal tritides and OBT was added to Section 5.2.4.

Added new information on sensitivity in 1950. In Table 5-10, a new 

footnote b on reporting level in 1964 was added. In Table 5-14, there 

was a change in MDA for the period 9/1964 through 1973. More 

explanatory text was added to the footnotes. Additional history of 233U 

exposure and bioassay method in 1970 was added to Section 5.2.5.

Updated impurity concentrations in the 233U and added new Tables 5-

16 and 5-17 on impurities. In Section 5.2.8, Information was added in 

regard to 147Pm urinalysis development in 1962. Added MDA 

information on urinalysis 1963-64. In vivo counting for 147Pm was 

mentioned. In Section 5.2.9, historical material was removed. New 

information was referenced in the Hanford Site Description. Added 

information on 1948 Po-210 urinalysis procedure and statement 

about monitoring in 1954. Text was added to Section 5.2.10 on purity 

of Np. Added information about urinalysis method in 1960. Text was 

added to Section 5.2.11 on thorium urinalysis method circa 1955.

Text was removed in Section 5.3.4 on historical exposure to thorium.

New information was referenced in the Hanford Site Description.

Buildings were added to Table 5-26. Mentioned early thorium 

urinalysis procedure circa 1955 and first whole body counting 1962-

63. New intake guidance for alpha-emitters in tank farm mixtures was 

added to Section 5.4.1. In Section 5.5.2, removed instructions to 

dose reconstructors concerning uncertainty. In Section 5.6, added 

reference to OTIB-0054 for fission product intakes. Corrected 

operating dates for the Fast Flux Test Facility. Mentioned that routine 

thyroid counts extended into the 1950s. In Table 5-30, added new 

intake values for tritium for 1955 and discussed reason in the text.

Added statements on the tritium bioassay analysis as discussed in 

Attachment C. Added guidance for unmonitored worker intakes of 

233U in Z Plant. Added new guidance for many of the specific 

buildings and added some new buildings to the list. Updated Table 5-

31 to reflect new guidance in the text. Updated Attributions and 

References sections to reflect changes in the text. Incorporates 

formal internal and NIOSH review comments. Provides additional 

guidance for assigning intakes for exotic radionuclides to 

unmonitored workers in Section 5.6. Constitutes a total rewrite of the 

document. Training required: As determined by the Objective 

Manager. Initiated by Donald E. Bihl and Fred L. Duncan.
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 5 of 180

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

10/20/2010 04 Revision to make corrections to GSDs and 95th percentile intake 

rates in Attachment C (coworker study) for all radionuclides 

(plutonium, uranium, strontium-90, promethium-137, zinc-65, sodium24, cesium-137). Incorporates formal internal and NIOSH review 

comments. Training required: As determined by the Objective 

Manager. Initiated by Matthew Arno and Fred L. Duncan.

11/05/2012 05 Revision to include NIOSH-proposed SEC-00201 regarding U-233, 

highly enriched uranium (HEU), neptunium, and thorium, from July 1, 

1972, through December 31, 1983. SEC statements added or 

changes made to Sections 5.1.2, 5.2.5, 5.2.9, 5.2.10, 5.2.11, 5.3.4, 

5.6.1, 5.6.2, and Table 5-31. Unmonitored worker intakes for thorium, 

neptunium, and U-233 deleted from Table 5-31 and associated text.

Corrected Table 5-17 to Th-228, and Attachment A number to 1.2 E11 (the minus sign was missing). Reference for TBD-6000 updated.

Incorporates formal internal and NIOSH review comments. Training 

required: As determined by the Objective Manager. Initiated by Fred 

L. Duncan. 

11/16/2015 06 Revision to include the SEC class determination for SEC-00226 due 

to inadequate bioassay monitoring for contractors and subcontractors 

of the DOE, except those of specified prime contractors, during the 

period from January 1, 1984, through December 31, 1990. SEC 

statements were added or changes made to Sections 5.1.2, 5.6, and 

5.6.2. Added titles and heading rows to some tables that did not have 

them. No information in those tables was changed. Incorporates 

formal internal and NIOSH review comments. Added Tables 5-1a

through 5-1e to Section 5.1.2 to summarize internal doses that can

and cannot be reconstructed for partial dose reconstructions during 

the SEC periods. Subsequent clarifications were made in Sections

5.6.1 and 5.6.2. Constitutes a total rewrite of the document. Training 

required: As determined by the Objective Manager. Initiated by Fred 

L. Duncan.

09/01/2020 07 Revision initiated to include adding the potential and intake rate 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 

Matthew G. Arno.

TRADEMARK INFORMATION

MATLAB® is a registered trademark of The Mathworks, Inc. in the United States and/or other 

countries.
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 6 of 180

 

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations................................................................................................................11

5.1 Introduction ...............................................................................................................................14

5.1.1 Purpose.........................................................................................................................15

5.1.2 Scope............................................................................................................................15

5.1.3 Special Exposure Cohort ..............................................................................................15

5.1.4 Background and Overview............................................................................................17

5.2 In Vitro Minimum Detectable Activities, Analytical Methods, and Reporting Protocols.............23

5.2.1 Plutonium ......................................................................................................................24

5.2.2 Americium .....................................................................................................................30

5.2.3 Curium...........................................................................................................................33

5.2.4 Tritium ...........................................................................................................................35

5.2.5 Uranium.........................................................................................................................36

5.2.6 Fission Products............................................................................................................44

5.2.7 Strontium.......................................................................................................................45

5.2.8 Promethium...................................................................................................................46

5.2.9 Polonium .......................................................................................................................47

5.2.10 Neptunium.....................................................................................................................49

5.2.11 Other Limited-Exposure Radionuclides.........................................................................49

5.3 In Vivo Minimum Detectable Activities, Analytical Methods, and Reporting Protocols .............51

5.3.1 Whole-Body Counters ...................................................................................................51

5.3.2 Chest Counters .............................................................................................................56

5.3.3 Thyroid Counters...........................................................................................................58

5.3.4 Thorium Exposure and Monitoring ................................................................................59

5.3.5 Head Counters and Other Counts.................................................................................62

5.3.6 General Notes about Items in the Database ................................................................. 63

5.4 Mixtures ....................................................................................................................................64

5.4.1 Fission and Activation Products ....................................................................................64

5.5 Interferences and Uncertainties ................................................................................................65

5.5.1 Contamination of Samples ............................................................................................65

5.5.2 Uncertainties .................................................................................................................66

5.6 Unmonitored Workers ...............................................................................................................67

5.6.1 Radionuclides Not Presented in Attachment C .............................................................67

5.6.2 Assignments for Unmonitored Workers ........................................................................72

5.7 Attributions and Annotations .....................................................................................................77

References ...........................................................................................................................................92

Glossary..............................................................................................................................................103

ATTACHMENT A TOLERANCE DOSE AND TOLERANCE VALUES ........................................108

ATTACHMENT B CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS ...............110
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 7 of 180

 

ATTACHMENT C INTERNAL DOSIMETRY CO-EXPOSURE DATA..........................................119

LIST OF TABLES 

TABLE TITLE PAGE

5-1a Summary of Hanford SEC impacts for plutonium .....................................................................18

5-1b Summary of Hanford SEC impacts for fission products............................................................19

5-1c Summary of Hanford SEC impacts for uranium........................................................................20

5-1d Summary of Hanford SEC impacts for americium ....................................................................21

5-1e Summary of Hanford SEC impacts for other radionuclides ......................................................22

5-2 Routine plutonium urinalysis detection levels ...........................................................................25

5-3 MDAs for nonroutine plutonium fecal analysis..........................................................................27

5-4 MDAs for nonroutine plutonium urinalysis ................................................................................27

5-5 Specific activities of Hanford reference weapons-grade plutonium mixture (6%).....................28

5-6 Activity ratios of Hanford reference weapons-grade plutonium mixture (6%)...........................28

5-7 Specific activities of Hanford reference fuel-grade plutonium mixture (12%) ...........................28

5-8 Activity ratios of Hanford reference fuel-grade plutonium mixture (12%)..................................28

5-9 Specific activities of Hanford reference commercial power fuel-grade plutonium mixture........29

5-10 Activity ratios of Hanford reference commercial power fuel-grade plutonium mixture ..............30

5-11 Routine 241Am urinalysis detection levels..................................................................................32

5-12 MDAs for nonroutine 241Am fecal analysis ................................................................................33

5-13 MDAs for nonroutine 241Am fecal urinalysis ..............................................................................33

5-14 Routine curium urinalysis detection levels ................................................................................34

5-15 MDAs for nonroutine curium fecal analysis...............................................................................34

5-16 MDAs for nonroutine curium urinalysis .....................................................................................34

5-17 Routine tritium urinalysis detection levels .................................................................................36

5-18 Radiological characteristics of Hanford RU ..............................................................................37

5-19 Radiological characteristics of Hanford commercial fuel ..........................................................37

5-20 Impurities in RU at Hanford.......................................................................................................37

5-21 Inhalation class for Hanford uranium compounds.....................................................................38

5-22 Routine uranium urinalysis detection levels, elemental ............................................................40

5-23 Routine uranium urinalysis detection levels, isotopic................................................................41

5-24 MDAs for nonroutine uranium fecal analysis ............................................................................41

5-25 MDAs for nonroutine uranium urinalysis ...................................................................................42

5-26 Highest level of impurities as measured for the three campaigns ............................................43

5-27 Highest impurity concentrations in 233U .....................................................................................43

5-28 Routine fission product urinalysis detection levels....................................................................44

5-29 Routine 90Sr urinalysis detection levels.....................................................................................46

5-30 Routine 147Pm urinalysis detection levels..................................................................................47

5-31 Routine 210Po urinalysis detection levels...................................................................................48

5-32 MDAs for nonroutine 210Po fecal analysis .................................................................................48

5-33 MDAs for nonroutine 210Po urinalysis........................................................................................49

5-34 Routine whole-body counting detection levels..........................................................................51

5-35 Mean body burdens of 137Cs from fallout in the United States..................................................56

5-36 Routine chest counting detection levels....................................................................................57

5-37 Buildings with thorium processing.............................................................................................60

5-38 232Thorium intakes from whole-body count measurements, type M..........................................61

5-39 232Thorium intakes from whole-body count measurements, type S ..........................................61

5-40 Ratios of 232Th to 228Ac in the whole body after end of 1- to 6-yr intake periods, type M ..........62

5-41 Ratios of 232Th to 228Ac in the whole body after end of 1- to 6-yr intake periods, type S...........63

5-42 Statistical parameters of Hanford recorded tritium doses and associated intakes ...................68
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5-43 Intakes for workers with external monitoring but essentially no bioassay.................................72

A-1 Example tolerance limits, 1945...............................................................................................108

B-1 Codes and radionuclides associated with bioassay at Hanford..............................................111

B-2 Sample type codes .................................................................................................................114

B-3 Bioassay reason codes...........................................................................................................114

B-4 Excreta sample kit codes ........................................................................................................115

B-5 In vivo body location codes.....................................................................................................116

B-6 Excreta unit codes ..................................................................................................................116

B-7 Excreta processing codes.......................................................................................................116

B-8 Excreta laboratory codes ........................................................................................................117

B-9 Excreta no-sample codes .......................................................................................................117

B-10 In vivo invalid result codes......................................................................................................117

B-11 INTERTRAC mode-of-intake codes........................................................................................117

B-12 INTERTRAC evaluation reason codes ...................................................................................118

B-13 INTERTRAC source-of-intake codes ......................................................................................118

B-14 INTERTRAC miscellaneous codes .........................................................................................118

B-15 Special whole-body count resolution codes............................................................................118

C-1 Plutonium reporting levels.......................................................................................................128

C-2 Time intervals for statistical analysis of plutonium urine samples...........................................128

C-3 Uranium reporting levels .........................................................................................................129

C-4 Time intervals for statistical analysis of uranium urine samples .............................................130

C-5 Strontium reporting levels .......................................................................................................130

C-6 Time intervals for statistical analysis of 90Sr urine samples ....................................................131

C-7 Promethium reporting levels ...................................................................................................132

C-8 Time intervals for statistical analysis of 147Pm urine samples .................................................132

C-9 Time intervals for statistical analysis of 65Zn whole-body counts............................................133

C-10 Time intervals for statistical analysis of 24Na whole-body counts ...........................................134

C-11 Time intervals for statistical analysis of 137Cs whole-body counts ..........................................134

C-12 Plutonium mixture usage ........................................................................................................138

C-13 Type M 239+240Pu intake periods and rates ..............................................................................139

C-14 Type S 239+240Pu intake period and rates.................................................................................140

C-15 Type SS 239+240Pu intake period and rates ..............................................................................140

C-16 Type F uranium intake periods and rates ...............................................................................141

C-17 Type M uranium intake periods and rates...............................................................................141

C-18 Type S uranium intake periods and rates ...............................................................................141

C-19 Type F 90Sr intake periods and rates ......................................................................................142

C-20 Type M 147Pm intake periods and rates ..................................................................................142

C-21 Type S 147Pm intake periods and rates ...................................................................................142

C-22 Type S 65Zn inhalation intake periods and rates .....................................................................143

C-23 Zinc-65 ingestion intake periods and rates .............................................................................143

C-24 Type F 24Na inhalation intake periods and rates.....................................................................143

C-25 Sodium-24 ingestion intake periods and rates........................................................................144

C-26 Type F 137Cs intake periods and rates ....................................................................................144

C-27 Type M 239Pu intake periods and rates ...................................................................................146

C-28 Type S 239Pu intake period and rates......................................................................................146

C-29 Type SS 239Pu intake period and rates ...................................................................................146

C-30 Uranium type F intake periods and rates ................................................................................146

C-31 Uranium type M intake periods and rates ...............................................................................146

C-32 Uranium type S intake periods and rates................................................................................147

C-33 Type F 90Sr intake periods and rates ......................................................................................147

C-34 Type M 147Pm intake periods and rates ..................................................................................148

C-35 Type S 147Pm intake periods and rates ...................................................................................148

C-36 Zinc-65 inhalation intake periods and rates ............................................................................149
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C-37 Zinc-65 ingestion intake periods and rates .............................................................................149

C-38 Type F 24Na inhalation intake periods and rates.....................................................................150

C-39 Type F 24Na ingestion intake periods and rates......................................................................150

C-40 Type F 137Cs intake periods and rates ....................................................................................150

C-41 Statistical summary of 239Pu 24-hour urinary excretion rates, 1946 to 1988...........................151

C-42 Statistical summary of uranium 24-hour urinary excretion rates, 1948 to 1988......................160

C-43 Statistical summary of 90Sr 24-hour urinary excretion rates, 1966 to 1988.............................166

C-44 Statistical summary of 147Pm 24-hour urinary excretion rates, 1967 to 1977..........................167

C-45 Statistical summary of 65Zn measured in whole-body counts, 1960 to 1984 ..........................169

C-46 Statistical summary of 24Na measured in whole-body counts, 1960 to 1984..........................174

C-47 Statistical summary of 137Cs measured in whole-body counts, 1960 to 1988.........................177

C-48 Hanford monthly reports .........................................................................................................179

C-49 Tritium hard-copy bioassay data.............................................................................................179

LIST OF FIGURES

FIGURE TITLE PAGE

C-1 Plutonium bioassay data in REX and from Hanford monthly reports......................................124

C-2 Uranium bioassay data in REX and from Hanford monthly reports ........................................125

C-3 Tritium bioassay data..............................................................................................................135

C-4 Individual tritium bioassay or dose data by year.....................................................................136

C-5 Tritium dose from hard-copy records and the main text .........................................................137

C-6 Production rates in tons of uranium times fuel burnup, 1944 to 1972.....................................145

C-7 50th-percentile plutonium urinalysis data for intakes of type M material, 1946 to 1948 .........154

C-8 50th-percentile plutonium urinalysis data for intakes of type M material, 1949 to 1952 .........155

C-9 50th-percentile plutonium urinalysis data for intakes of type M material, 1953 to 1981 .........155

C-10 50th-percentile plutonium urinalysis data for intakes of type M material, 1982 to 1988 .........155

C-11 Predicted 50th-percentile urinary excretion of type M plutonium from 1946 to 1988 based 

on four independent intakes, compared to bioassay data ......................................................156

C-12 50th-percentile plutonium urinalysis data for intakes of type S material, 1946 to 1948 ..........156

C-13 50th-percentile plutonium urinalysis data for intakes of type S material, 1949 to 1952 ..........156

C-14 50th-percentile plutonium urinalysis data for intakes of type S material, 1953 to 1981 ..........157

C-15 50th-percentile plutonium urinalysis data for intakes of type S material, 1982 to 1988 ..........157

C-16 Predicted 50th-percentile urinary excretion of type S plutonium from 1946 to 1988 based 

on four independent intakes, compared to bioassay data ......................................................157

C-17 50th-percentile plutonium urinalysis data for intakes of type SS material, 1946 to 1948........158

C-18 50th-percentile plutonium urinalysis data for intakes of type SS material, 1949 to 1952........158

C-19 50th-percentile plutonium urinalysis data for intakes of type SS material, 1953 to 1981........159

C-20 50th-percentile plutonium urinalysis data for intakes of type SS material, 1982 to 1988........159

C-21 Predicted 50th-percentile urinary excretion of type SS plutonium from 1946 to 1988 

based on four independent intakes, compared to bioassay data............................................160

C-22 50th-percentile uranium urinalysis data for intakes of type F material, 1948 to 1988.............163

C-23 50th-percentile uranium urinalysis data for intakes of type M material, 1948 to 1988 ............163

C-24 50th-percentile uranium urinalysis data for intakes of type S material, 1948 to 1952.............163

C-25 50th-percentile uranium urinalysis data for intakes of type S material, 1953 to 1956.............164

C-26 50th-percentile uranium urinalysis data for intakes of type S material, 1957 to 1961.............164

C-27 50th-percentile uranium urinalysis data for intakes of type S material, 1962 to 1974.............164

C-28 50th-percentile uranium urinalysis data for intakes of type S material, 1975 to 1983.............165

C-29 50th-percentile uranium urinalysis data for intakes of type S material, 1984 to 1988.............165

C-30 Predicted 50th-percentile urinary excretion of type S uranium from 1946 to 1988 based 

on six independent intakes, compared to bioassay data ........................................................165
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C-31 50th-percentile strontium urinalysis data for intakes of type F material, 1965 to 1988 ...........167

C-32 50th-percentile promethium urinalysis data for intakes of type M material, 1966 to 1969 ......167

C-33 50th-percentile promethium urinalysis data for intakes of type M material, 1970 to 1979 ......168

C-34 Predicted 50th-percentile urinary excretion of type M promethium from 1966 to 1979 

based on two independent intakes, compared to bioassay data ............................................168

C-35 50th-percentile promethium urinalysis data for intakes of type S material, 1966 to 1969.......168

C-36 50th-percentile promethium urinalysis data for intakes of type S material, 1970 to 1979.......169

C-37 Predicted 50th-percentile urinary excretion of type S promethium from 1966 to 1979 

based on two independent intakes, compared to bioassay data ............................................169

C-38 50th-percentile zinc whole-body counting data for inhalation intakes of type S material, 

1960 to 1963...........................................................................................................................171

C-39 50th-percentile zinc whole-body counting data for inhalation intakes of type S material, 

1964 to 1967...........................................................................................................................172

C-40 50th-percentile zinc whole-body counting data for inhalation intakes of type S material, 

1968 to 1970...........................................................................................................................172

C-41 50th-percentile zinc whole-body counting data for inhalation intakes of type S material, 

1971 to 1972...........................................................................................................................172

C-42 50th-percentile zinc whole-body counting data for inhalation intakes of type S material, 

1973 to 1977...........................................................................................................................173

C-43 50th-percentile zinc whole-body counting data for inhalation intakes of type S material, 

1978 to 1984...........................................................................................................................173

C-44 Predicted 50th-percentile urinary excretion of inhaled type S zinc from 1960 to 1984 

based on six independent chronic intakes and one acute intake, compared to bioassay 

data.........................................................................................................................................173

C-45 50th-percentile zinc whole-body counting data for ingestion intakes, 1960 to 1984...............174

C-46 50th-percentile sodium whole-body counting data for inhalation intakes of type F 

material, 1960 to 1984 ............................................................................................................176

C-47 50th-percentile sodium whole-body counting data for ingestion intakes, 1960 to 1984..........176

C-48 50th-percentile 137Cs whole-body counting data for inhalation intakes of type F material, 

1960 to 1988...........................................................................................................................179
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AWE atomic weapons employer

Bq becquerel

Ci curie

d day

D&D decontamination and decommissioning

DAC derived air concentration

DDCP dibutyl-N,N-diethylcarbamylphosphonate

DIL derived investigation level

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

DTPA diethylene triamine pentaacetic acid

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

F fast (absorption type)

FFTF Fast Flux Test Facility

g gram

GeLi lithium-drifted germanium

GM geometric mean

GSD geometric standard deviation

HERB Health-Related Energy Research Branch

HEU highly enriched uranium

HIE Hanford Internal Exposure (database)

HPGe high-purity germanium

hr hour

HT elemental tritium (tritiated gas)

HTO tritiated water vapor

ICRP International Commission on Radiological Protection

IDOT Internal Dosimetry Tool

IMBA Integrated Modules for Bioassay Assessment

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt

kg kilogram

KPA kinetic phosphorescence analysis

L liter

lb pound

LANL Los Alamos National Laboratory

LEPD low-energy photon detector (also a computer code to indicate use of the LEPD)

LSL-II Life Sciences Laboratory-II
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m meter

M medium (absorption type)

MDA minimum detectable amount

MDC minimum detectable concentration

MeV megaelectron-volt

min minute

mg milligram

mL milliliter

MPBB maximum permissible body burden

MPC maximum permissible concentration

mR milliroentgen

mrem millirem

mrep millirep

mSv millisievert

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NU natural uranium

OBT organically bound tritium

ORAU Oak Ridge Associated Universities

ORE Occupational Radiological Exposure (database)

pCi picocurie

PER program evaluation report

PFP Plutonium Finishing Plant

PNNL Pacific Northwest National Laboratory

PoBe polonium-beryllium (alloy)

ppb parts per billion

ppm parts per million

PRTR Plutonium Recycle Test Reactor

PUREX Plutonium-Uranium Extraction (Plant)

RDA reliably detectable activity

REDOX Reduction-Oxidation (Plant)

REX Radiological Exposure (database)

RU recycled uranium

S slow (absorption type)

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

SS super S (absorption type)

TBD technical basis document

TPU total propagated uncertainty

TTA thenoyl trifluoroacetone

USC United States Code

UST United States Testing Company

wk week

WLM working level month
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yr year

μCi microcurie

μg microgram

μm micrometer
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5.1 INTRODUCTION

Technical basis documents (TBDs) and site profile documents are not official determinations made by 

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

documents that provide 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 United 

States Code (USC) 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 USC 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) describes the internal dosimetry program at the Hanford Site 

and provides the technical basis to be used to evaluate the internal occupational radiation dose for 

EEOICPA claims.

5.1.2 Scope

Hanford operations have played an important role in the development of the U.S. nuclear weapons 

program. This TBD is part of the overall Hanford site profile, which describes plant facilities and 

processes, historical information, and environmental data in relation to dose reconstruction for 

Hanford workers. It contains supporting documentation to assist in the reconstruction of occupational 

internal doses from these activities.

The methods and concepts of measuring occupational internal doses to workers have evolved since 

the beginning of operations at the Hanford Site. An objective of this document is to provide supporting 

technical data to evaluate internal occupational doses that can reasonably be associated with worker 

radiation exposures. In addition, this document presents the technical basis of methods used to 

prepare Hanford worker dose information for input to the Interactive RadioEpidemiological Program 

(IREP). 

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. Attachment C, 

Internal Dosimetry Co-Exposure Data, provides guidance on assigning intakes for the years before 

bioassay.

5.1.3 Special Exposure Cohort

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

of Hanford employees as additions to the SEC based on the findings and recommendations of NIOSH 

and the Advisory Board on Radiation and Worker Health:

October 1, 1943, through August 31, 1946

Employees of the Department of Energy (DOE), its predecessor agencies, or DOE 

contractors or subcontractors who were monitored or should have been monitored for 

internal radiological exposures while working at the Hanford Engineer Works in: the 300 

Area fuel fabrication and research facilities from October 1, 1943, through August 31, 

1946; the 200 Area plutonium separation facilities from November 1, 1944 through 

August 31, 1946; or the 100 B, D, and F reactor areas from September 1, 1944, through 

August 31, 1946; for a number of work days aggregating at least 250 work days or in 

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

classes of employees in the Special Exposure Cohort [Leavitt 2007].

September 1, 1946, through December 31, 1968

All employees of the Department of Energy (DOE), its predecessor agencies, and DOE 

contractors or subcontractors who worked from: 

1. September 1, 1946 through December 31, 1961 in the 300 area; or

2. January 1, 1949 through December 31, 1968 in the 200 areas (East and West)

at the Hanford Nuclear Reservation in Richland, Washington, for a number of work days 

aggregating at least 250 work days occurring either solely under this employment or in 
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combination with work days within the parameters established for one or more other 

classes of employees in the Special Exposure Cohort [Leavitt 2008].

October 1, 1943, through June 30, 1972

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

and subcontractors who worked at the Hanford site in Richland, Washington, from 

October 1, 1943 through June 30, 1972, 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 [Sebelius 2009].

July 1, 1972, through December 31, 1983

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

contractors and subcontractors who worked at the Hanford Engineer Works in Richland, 

Washington, from July 1, 1972, 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 [Sebelius 2012].

January 1, 1984, through December 31, 1990

All employees of Department of Energy contractors and subcontractors (excluding 

employees of the following Hanford prime contractors during the specified time periods: 

Battelle Memorial Institute, January 1, 1984, through December 31, 1990; Rockwell 

Hanford Operations, January 1, 1984, through June 28, 1987; Boeing Computer 

Services Richland, January 1, 1984, through June 28, 1987; UNC Nuclear Industries,

January 1, 1984, through June 28, 1987; Westinghouse Hanford Company, January 1, 

1984, through December 31, 1990; and Hanford Environmental Health Foundation, 

January 1, 1984, through December 31, 1990), who worked at the Hanford site in 

Richland, Washington, during the period from January 1, 1984, through December 31, 

1990, 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 [Burwell 2015].

Dose reconstruction guidance in this document is presented to provide a technical basis for partial 

dose reconstructions for claims not covered within the SEC classes through December 31, 1983; 

partial dose reconstructions for nonexcepted contractor and subcontractor claims not covered within 

the SEC class from January 1, 1984, through December 31, 1990; and full dose reconstructions for 

DOE employees and employees of excepted prime contractors from January 1, 1984, through 

December 31, 1990. NIOSH found that it was not feasible to reconstruct doses for inadequately 

monitored radionuclides such as polonium, thorium, and neptunium through June 30, 1972 [NIOSH 

2009], and for highly enriched uranium (HEU), 233U, thorium, or neptunium, from July 1, 1972, through 

December 31, 1983 [NIOSH 2012]. In addition, NIOSH found that it was infeasible to reconstruct 

doses for employees of DOE contractors and subcontractors, except for those of employers excluded 

above, during the period from January 1, 1984, through December 31, 1990, due to inadequate 

bioassay monitoring [Burwell 2015a]. 

Although NIOSH found that it is not possible to completely reconstruct radiation doses for the added 

classes, NIOSH intends to use any internal and external monitoring data that may become available 

for an individual claim (and that can be interpreted using existing NIOSH dose reconstruction 

processes or procedures). Therefore, dose reconstructions for individuals who were employed at the 

Hanford site during the period from October 1, 1943, through December 31, 1990, but who do not 

qualify for inclusion in the SEC, may be performed using these data as appropriate. The co-exposure
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intake rates in Attachment C are based on bioassay data that do not adequately represent the added 

SEC class for the period from January 1, 1984, through December 31, 1990; therefore, co-exposure

intake rates in Attachment C will not be assigned for that period for partial dose reconstructions for 

claims not covered within the SEC. Tables 5-1a through 5-1e summarize internal doses that can, and 

cannot, be reconstructed for partial dose reconstructions during the SEC periods. The tables 

respectively address plutonium, fission products, uranium, americium, and other radionuclides. More 

detailed information can be found in the sections for each radionuclide and in Section 5.6 for 

unmonitored workers. 

5.1.4 Background and Overview

When the first reactor became operational on the Hanford Site, there were no programs to monitor an 

employee for internal dose with the exception of measuring particles in the air. The site operated a 

fuel manufacturing facility, three reactors, and four processing plants from 1943 to 1946 before a 

bioassay program was in place. The responsibility for personnel monitoring was with the Medical 

Department. 

In the 1940s, the radiation protection community used the term “tolerance” to describe dose limits and 

other values such as air concentration values. This was before the National Council on Radiation 

Protection and Measurements (NCRP) introduced the terms maximum permissible body burden 

(MPBB) and maximum permissible concentration (MPC). Air concentration tolerance levels were 

based on dose rates to significantly impacted organs. Examples at Hanford were 0.01 rep/d from 

alpha emitters, 1 rep/d to the thyroid from beta and gamma emitters, or total accumulation in the body 

of a radionuclide with comparable effects to 0.1 μg of radium. The latter limit was used to establish the 

limit of 0.5 μg for total accumulation of plutonium in the body. The radiation protection program was 

directed toward preventing workers from acquiring a tolerance dose or being exposed to tolerance air 

concentrations. Workers were supposed to wear respiratory protection if an air concentration 

exceeded the tolerance level or if a tolerance level was anticipated to be exceeded for a given job. In 

some cases, a lower air concentration was established as the respirator-required air concentration.

The tolerance air concentrations in 1945 were 5 × 10-10 μg/cm3 (4 × 10-11 μCi/cm3) for product 

(meaning plutonium), 1 × 10-8 μCi/cm3 for fission and activation products, and 1.5 × 10-4 μg/cm3

(1.1 × 10-10 μCi/cm3) for uranium [Cantril 1945]. By 1947, the respirator-required air concentrations 

were 2 × 10-11 μg/cm3 (2 × 10-12 μCi/cm3) for plutonium, 1 × 10-9 μCi/cm3 for fission products (assumed 

to mean particulate beta emitters), and 5 × 10-5 μg/cm3 (3.5 × 10-11 μCi/cm3) for uranium [Parker 1947; 

Patterson 1949]. Additional discussion on tolerance levels is provided in Attachment A.

According to the history by Wilson [1987], one of the priority tasks for a special studies group that was 

formed in 1944 was to determine a way to measure plutonium in the body. Limits on the amount of 

plutonium in the body were set as early as 1944 and, after experimentation with various methods, 

routine urine sampling and analysis for plutonium were initiated in 1946. Urinalysis for uranium seems 

also to have started in 1946 and was well established by 1948. Urinalysis for fission products started 

in this period as well, although the Wilson document indicates that separation from 40K was not always 

successful before 1949. Since then, monitoring for numerous radionuclides has occurred at Hanford 

because of the complex scope of work over the years, the many research projects, special 

campaigns, and other activities. In addition, numerous techniques have been used because of 

improvements in techniques. The major sources of intakes have been plutonium, 241Am either as an 

ingrown contaminant in the plutonium or as a separated waste product, uranium, fission products, 

activation products, and tritium. However, the records list a wide spectrum of radionuclides that were 

monitored and an even longer list of codes that identify the radionuclides, groups of radionuclides, 

specific measurement techniques, or combinations of radionuclides and techniques. Many of the 

radionuclides apply to a small set of workers on a research project or to workers (for instance, 

radiation monitoring technicians) whose tasks might have exposed them to many different sources.
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Table 5-1a. Summary of Hanford SEC impacts for plutonium.

Applicable 

period

Class 

description

Doses that cannot

be reconstructed

Doses that can 

be reconstructed Notes

10/01/1943–

08/31/1946

All employees Doses from plutonium cannot 

be reconstructed because of 

inadequate monitoring.

Plutonium co-exposure intake 

rates in Attachment C cannot 

be assigned.

Individuals who have adequate and valid bioassay results 

for plutonium in their records (not expected) can have 

dose reconstructed for this period. 

Doses from environmental intakes of plutonium 

(calculated from emissions source terms and atmospheric 

dispersion modeling) can be assigned per Note 1 to table 

7-3 in the Evaluation Report.

SEC-0057-1, later 

expanded to all 

employees by SEC0152

09/01/1946–

06/30/1972

All employees Not applicable; all doses can 

be reconstructed

Plutonium doses can be assigned for adequately 

monitored workers. Plutonium co-exposure intake rates in 

Attachment C can be assigned for unmonitored or 

inadequately monitored workers

SEC-0057-1 and SEC0152 did not list 

plutonium as infeasible

07/01/1972–

12/31/1983

All employees Not applicable; all doses can 

be reconstructed

Plutonium doses can be assigned for adequately 

monitored workers. Plutonium co-exposure intake rates in 

Attachment C can be assigned for unmonitored or 

inadequately monitored workers.

SEC-0201 did not list 

plutonium as infeasible

01/01/1984–

12/31/1990

All employees of 

Department of 

Energy 

contractors and 

subcontractors 

excluding 

employees of 

listed Hanford 

prime 

contractors.

For workers included in the 

SEC, the co-exposure intake 

rates for unmonitored or 

inadequately monitored 

workers in Section 5.6 and 

Attachment C are based on 

bioassay data that do not 

adequately represent the 

added SEC class and will not 

be assigned for that period.

Plutonium doses can be assigned using internal

monitoring data that may become available for an 

individual claim in the SEC class, and can be assigned for 

employees of the following Hanford prime contractors 

during the specified periods; and co-exposure intake 

rates in Attachment C can be assigned for employees of 

the following Hanford prime contractors during the 

specified periods:

Rockwell Hanford Operations, Boeing Computer Services 

Richland, and UNC Nuclear Industries, January 1, 1984 

through June 28, 1987; and Battelle Memorial Institute, 

Westinghouse Hanford Company, and Hanford 

Environmental Health Foundation, January 1, 1984 

through December 31, 1990. 

SEC-0226
 

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Table 5-1b. Summary of Hanford SEC impacts for fission products.

Applicable 

period

Class 

description

Doses that cannot

be reconstructed

Doses that can 

be reconstructed Notes

10/01/1943–

08/31/1946

All employees Doses from fission products 

cannot be reconstructed 

because of inadequate 

monitoring.

Fission product co-exposure

intake rates in Section 5.6.2

(H-3 and I-131) and 

Attachment C (Sr-90 and Cs137) cannot be assigned.

Individuals who have adequate and valid bioassay results 

for fission products in their records can have dose 

reconstructed for this period (none found to date). 

Doses from environmental intakes of fission products 

(calculated from emissions source terms and atmospheric 

dispersion modeling) can be assigned per Note 1 to table 

7-3 in the Evaluation Report.

SEC-0057-1, later 

expanded to all 

employees by SEC0152

09/01/1946–

06/30/1972

All employees Not applicable; all doses can 

be reconstructed

Fission products doses can be assigned for adequately 

monitored workers. Fission products co-exposure intake 

rates in Attachment C can be assigned for unmonitored or 

inadequately monitored workers.

SEC-0057-2 and SEC0152 did not list fission 

products as infeasible

07/01/1972–

12/31/1983

All employees Not applicable; all doses can 

be reconstructed

Fission products doses can be assigned for adequately 

monitored workers. Fission products co-exposure intake 

rates in Attachment C can be assigned for unmonitored or 

inadequately monitored workers.

SEC-0201 did not list 

fission products as 

infeasible

01/01/1984–

12/31/1990

All employees of 

DOE contractors 

and 

subcontractors 

excluding 

employees of 

listed Hanford 

prime 

contractors.

For workers included in the 

SEC, the co-exposure intake 

rates for unmonitored or 

inadequately monitored 

workers in Section 5.6 and 

Attachment C are based on 

bioassay data that do not 

adequately represent the 

added SEC class and will not 

be assigned for that period.

Fission products doses can be assigned using internal

monitoring data that may become available for an 

individual claim in the SEC class, and can be assigned for 

employees of the following Hanford prime contractors 

during the specified periods; and co-exposure intake 

rates in Attachment C can be assigned for employees of 

the following Hanford prime contractors during the 

specified periods: 

Rockwell Hanford Operations, Boeing Computer Services 

Richland, and UNC Nuclear Industries, January 1, 1984 

through June 28, 1987; and Battelle Memorial Institute, 

Westinghouse Hanford Company, and Hanford 

Environmental Health Foundation, January 1, 1984 

through December 31, 1990.

SEC-0226
 

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Table 5-1c. Summary of Hanford SEC impacts for uranium.

Applicable 

period

Class 

description

Doses that cannot

be reconstructed

Doses that can 

be reconstructed Notes

10/01/1943–

08/31/1946

All employees Not applicable; all doses can 

be reconstructed

SEC-0057-1 found that dose reconstruction for uranium 

was feasible. The uranium urinalysis program before 1949 

was unreliable; therefore, intakes for uranium should be 

assigned from December 1, 1943 (when machining of fuel 

rods began) through December 31, 1948 for workers in 

the 300 fuel fabrication areas in accordance with 

Attachment C and the Process/Job Title listed as 

Machining/Operator as presented in Battelle-TBD-6000, 

Site Profiles for Atomic Weapons Employers That Worked 

Uranium Metals [BT 2011].

SEC-0057-1, later 

expanded to all 

employees by SEC0152

09/01/1946–

06/30/1972

All employees Not applicable; all doses can 

be reconstructed

Uranium doses can be assigned for adequately monitored 

workers. Uranium co-exposure intake rates in Attachment 

C can be assigned for unmonitored or inadequately 

monitored workers (see statement above for assignments 

through December 31, 1948).

SEC-0057-2 and SEC0152 did not list 

uranium as infeasible

07/01/1972–

12/31/1983

All employees Not applicable; all doses can 

be reconstructed

Uranium doses can be assigned for adequately monitored 

workers. Uranium co-exposure intake rates in Attachment 

C can be assigned for unmonitored or inadequately 

monitored workers.

SEC-0201 did not list 

uranium as infeasible

01/01/1984–

12/31/1990

All employees of 

DOE contractors 

and 

subcontractors 

excluding 

employees of 

listed Hanford 

prime 

contractors.

For workers included in the 

SEC, the co-exposure intake 

rates for unmonitored or 

inadequately monitored 

workers in Section 5.6 and 

Attachment C are based on 

bioassay data that do not 

adequately represent the 

added SEC class and will not 

be assigned for that period.

Uranium doses can be assigned using internal monitoring 

data that may become available for an individual claim in 

the SEC class, and can be assigned for employees of the 

following Hanford prime contractors during the specified 

periods; and co-exposure intake rates in Attachment C 

can be assigned for employees of the following Hanford 

prime contractors during the specified periods: 

Rockwell Hanford Operations, Boeing Computer Services 

Richland, and UNC Nuclear Industries, January 1, 1984 

through June 28, 1987; and Battelle Memorial Institute, 

Westinghouse Hanford Company, and Hanford 

Environmental Health Foundation, January 1, 1984 

through December 31, 1990. 

SEC-0226
 

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Table 5-1d. Summary of Hanford SEC impacts for americium.

Applicable 

period

Class 

description

Doses that cannot

be reconstructed

Doses that can 

be reconstructed Notes

10/01/1943–

08/31/1946

All employees Plutonium co-exposure intake 

rates in Attachment C cannot 

be assigned; therefore, 

americium associated with 

plutonium co-exposure

cannot be assigned.

Americium separations from 

plutonium began in 1949, so 

it was not present in 1943-

1946.

Doses can be assigned for americium which is assessed 

from adequate and valid plutonium bioassay and included 

as part of the plutonium matrix (none found to date).

SEC-0057-1, later 

expanded to all 

employees by SEC0152

09/01/1946–

06/30/1972

All employees as 

expanded by 

SEC-0152

Doses from americium 

separated from plutonium 

cannot be reconstructed 

because of inadequate 

monitoring. Co-exposure

intake rates do not exist for 

separated americium, so it 

cannot be assigned.

Doses can be assigned for americium which is assessed 

from plutonium bioassay and included as part of the 

plutonium matrix.

From January 1, 1969 through June 30, 1972, doses from

separated americium can be assigned when there are 

adequate isotopic americium bioassay results.

SEC-0057-2 listed

separated americium 

as infeasible for 

September 1, 1946 

through December 31, 

1968 and SEC-0152 

expanded this to all 

employees due to 

undocumented worker 

movement

07/01/1972–

12/31/1983

All employees Not applicable; all doses can 

be reconstructed

Doses from separated americium can be assigned when 

there are adequate isotopic americium bioassay results. 

Doses can be assigned for americium which is assessed 

from plutonium bioassay and included as part of the 

plutonium matrix.

SEC-0201 did not list 

americium as 

infeasible

01/01/1984–

12/31/1990

All employees of 

DOE contractors 

and 

subcontractors 

excluding 

employees of 

listed Hanford 

prime 

contractors.

For workers included in the 

SEC, the co-exposure intake 

rates for unmonitored or 

inadequately monitored 

workers in Section 5.6 and 

Attachment C are based on 

bioassay data that do not 

adequately represent the 

added SEC class and will not 

be assigned for that period.

Americium doses associated with plutonium (based on 

bioassay or co-exposure) or isotopic americium can be 

assigned using internal monitoring data that may become 

available for an individual claim in the SEC class, and can 

be assigned for employees of the following Hanford prime 

contractors during the specified periods: 

Rockwell Hanford Operations, Boeing Computer Services 

Richland, and UNC Nuclear Industries, January 1, 1984 

through June 28, 1987; and Battelle Memorial Institute, 

Westinghouse Hanford Company, and Hanford 

Environmental Health Foundation, January 1, 1984 

through December 31, 1990. 

SEC-0226
 

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Table 5-1e. Summary of Hanford SEC impacts for other radionuclides.

Applicable 

period

Class 

description

Doses that cannot

be reconstructed

Doses that can 

be reconstructed Notes

10/01/1943–

06/30/1972

All employees SEC-0152 identified polonium, 

thorium, and neptunium exposures in 

specified areas from January 1, 1945 

through June 30, 1972 as infeasible 

for dose reconstruction due to 

inadequate monitoring. 

There are no co-exposure intakes 

available for polonium, thorium, 

neptunium, HEU, and U-233 to 

assign. There are also no coexposure intakes available for 

Cm-244 and Tc-99.

Doses may be reconstructed for HEU, U-233 and 

other radionuclides for which individual 

radionuclide-specific and valid bioassay data are 

available (very unlikely). 

Co-exposure data may be assigned for this 

period per Tables 5-1a through 5-1d (coexposure models are listed in Attachment C for 

plutonium, uranium, Sr-90, Cs-137, Pm-147, Na24, and Zn-65).

SEC-0152 expanded 

the SEC -0057-1 and 

0057-2 classes to 

include all employees 

due to lack of data of 

worker movement 

between areas

07/01/1972–

12/31/1983

All employees SEC -0201 identified HEU, U-233, 

thorium, and neptunium as infeasible 

for dose reconstruction due to 

inadequate monitoring. 

There are no co-exposure intakes 

available for polonium, thorium, 

neptunium, HEU, and U-233 to 

assign. 

Doses from HEU, U-233, thorium, neptunium, 

and other radionuclides may only be 

reconstructed for individuals with radionuclidespecific and valid bioassay data during a period 

that the radionuclide was identified by the SEC. 

Co-exposure data may be assigned for this 

period (co-exposure models are listed in 

Attachment C for plutonium, uranium, Sr-90, Cs137, Pm-147, Na-24, and Zn-65).

SEC-0201 

01/01/1984–

12/31/1990

All employees of 

DOE contractors 

and 

subcontractors 

excluding 

employees of 

DOE and listed 

Hanford prime 

contractors.

For workers included in the SEC0226, the co-exposure intake rates for 

unmonitored or inadequately 

monitored workers in Section 5.6 and 

Attachment C are based on bioassay 

data that do not adequately represent 

the added SEC class and will not be 

assigned for that period.

Doses can be assigned using internal monitoring 

data that may become available for an individual 

claim in the SEC class, and can be assigned for 

employees of the following Hanford prime 

contractors during the specified periods: 

Rockwell Hanford Operations, Boeing Computer 

Services Richland, and UNC Nuclear Industries, 

January 1, 1984 through June 28, 1987; and 

Battelle Memorial Institute, Westinghouse 

Hanford Company, and Hanford Environmental 

Health Foundation, January 1, 1984 through 

December 31, 1990. 

SEC-0226
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Attachment B provides a fairly exhaustive list of codes for analyses that dose reconstructors might 

encounter in the bioassay or internal dosimetry records for Hanford workers. Some of the codes were 

used for scheduling bioassay but not for reporting the results. For instance, IPA is a code for 

performing plutonium and americium separation chemistry and alpha spectrometry on an excreta 

sample, but the results would normally be reported separately for 238Pu, 239Pu, and 241Am. However, if 

the sample was not obtained or the results could not be reported due to analysis problems, the record 

shows just the IPA code with a reason for not obtaining a result [1]. Other codes refer to a type of in 

vivo count or a special type of sample analysis. For instance, LEPD is the code for performing an 

X- and gamma-ray analysis on an excreta sample using the low-energy photon detector (LEPD; a 

thin-window germanium detector). However, if anything was detected, the actual radionuclide was 

reported [2]. The code GOK shows on in vivo count hard-copy records during the 1960s and 1970s.

This refers to net counts per minute from an undetermined source in a low-energy region of the 

spectrum from NaI-based whole-body counters [3].

Other bioassay codes have been used to indicate the following:

• Sample type,

• In vivo count body location,

• Reason for the sample or count,

• Type of kit and some details about the sampling protocol,

• Laboratory used,

• Laboratory turnaround time versus analytical sensitivity,

• Units associated with the result, and

• Reason for not obtaining a valid excreta result or in vivo count.

In addition, there are codes that pertain to the nature of the intake:

• Reason for an intake assignment,

• Source of intake (at Hanford or another site),

• Nature of intake, and

• Mode of intake.

Tables that list and explain these codes are provided in Attachment B.

Attachment C, Internal Dosimetry Co-Exposure Data, provides guidance on assigning intakes for the 

years before bioassay.

5.2 IN VITRO MINIMUM DETECTABLE ACTIVITIES, ANALYTICAL METHODS, AND 

REPORTING PROTOCOLS

Most urinalysis records have, at some time, been entered into one or more databases. However, 

there are quite a few records missing from the databases, especially for some of the earliest urinalysis 

records including a significant number of the mixed fission product urinalysis records. In addition, 

there are examples of duplicate records in the databases. For any case where urinalysis might have 

been obtained before 1974, the hard-copy file for the case should be thoroughly reviewed for 

urinalysis results that might be missing from the database. The Hanford Internal Exposure (HIE) 

database was implemented in 1974 and was followed by the Occupational Radiological Exposure 

(ORE) database in 1983 and the Radiological Exposure (REX) database in 1993. In principle, the 

REX database has all the information from the previous databases but, as stated above, there are 

situations where some data were never entered in a database or did not get transferred from one 

database to another [4].
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The results from around 1946 to 1950 contain another anomaly, urinalysis records with no result and 

no volume. This might indicate that the sample was not turned in or the analysis failed, but experience 

has shown that this convention was also used to indicate a nondetection. In many cases the actual 

laboratory urinalysis results card is available in the worker’s file and would show if the analysis was 

performed but the results were below detection or not [5].

Home sampling began very early in the program (1946) [Wilson 1987], and has continued throughout 

the history of Hanford. Home sampling was used to prevent contamination of samples in the 

workplace. The sampling protocol that was used most frequently (kit code 1) was a simulated 24-hour 

sample that was obtained by sampling from evening through morning on two consecutive nights (or 

equivalent for shift workers). Termination samples were often simulated 12-hour samples (kit code 2) 

[6]. The dose reconstructor should look at the kit code to determine if sample results should be treated 

as a day’s excretion or adjusted to represent the daily excretion.

In vitro analyses were performed in house until the breakup of the main Hanford contractor (General 

Electric Company) in 1965. At that time, the U.S. Atomic Energy Commission (AEC) Richland 

Operations Office established a contract for in vitro analyses with the United States Testing Company 

(UST), which built a commercial low-level radiochemistry laboratory in north Richland and operated it 

until 1990. The responsibility for awarding and overseeing the contract was subsequently transferred 

to Battelle Memorial Institute (Battelle) as operators of the Pacific Northwest Laboratory, now called 

Pacific Northwest National Laboratory (PNNL). Until 2005, with the exception of a period between 

1990 and 1992, in vitro analyses were performed in the same facility. However, due to buyouts and 

mergers, the name of the laboratory changed in the following sequence: UST, International 

Technology Analytical Services, Quanterra Environmental Services, and Severn Trent Laboratories.

The current in vitro laboratory is General Engineering Laboratory [7].

Battelle defaulted on the contract with UST in June 1990, after which routine samples were collected 

and frozen [Lyon et al. 1991, 1992]. Between September and November 1990, temporary contracts 

and agreements were established and samples were analyzed at the following laboratories: Los 

Alamos National Laboratory (LANL; plutonium), TMA-Norcal (strontium), PNL-Analytical Chemistry 

Laboratory (325 Building; tritium), and Westinghouse Hanford Company (222-S Building; elemental 

uranium). In February 1991, International Technology Analytical Services began analyses for 

plutonium, americium, curium, and isotopic uranium. LANL was replaced by Oak Ridge National 

Laboratory and Reynolds Electrical and Engineering Company at the Nevada Test Site (plutonium) in 

April 1991. The contract with International Technology Analytical Services replaced the contract with 

UST, but the other laboratories continued to process samples until the backlog was worked off.

Therefore, the work at the temporary laboratories was finishing during late 1991 through early 1992; 

Hanford received the last results in March 1992. This interruption in the normal process of bioassay 

monitoring had two effects that show in the bioassay records for this period: (1) changes in the 

minimum detectable amount (MDAs) for the various analyses reflecting capabilities at the interim 

laboratories; and (2) delays in collection, analysis, and followup to high routine results from the normal 

pattern [Lyon et al. 1991, 1992]. 

5.2.1 Plutonium

As presented in Section 5.1.3, an SEC class was designated in which NIOSH found that it was 

infeasible to reconstruct doses for plutonium before August 31, 1946 due to inadequate monitoring. 

Therefore, dose reconstructions will not include any doses from plutonium before August 31, 1946.

By far the most serious intakes at Hanford involved plutonium and 241Am. Routine urinalyses for 

plutonium started in September 1946 [Wilson 1987]. The first plutonium bioassay analysis consisted 

of lanthanum fluoride precipitation and thenoyl trifluoroacetone (TTA) extraction and gross alpha 

counting. Electrodeposition on a stainless-steel disk in combination with nuclear track emulsion 
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(autoradiography) started in December 1952. Detection levels for these and subsequent procedures 

are listed in Table 5-2. The definition of “detection level” no doubt changed over the years, but the 

levels in Table 5-2 fit reasonably with the concept of limit of detection or MDA. For example, Wilson 

[1987] states: 

From statistical evaluations of data collected in 1953, the true detection limit with nucleartrack film was determined. These evaluations showed 0.05 dpm was achievable within 

reasonable confidence levels. Occasionally recovery, counting, etc., allowed detection 

levels to be as low as 0.028 dpm and for a short period, a level of 0.027 dpm was reached 

and used as the detection level ... This practice [of recording lower detection levels] was 

discontinued and the more conservative 0.05 dpm was used routinely even though lower 

levels were possible part of the time.

Table 5-2. Routine plutonium urinalysis detection levels (dpm/sample).

a

Period MDA Decision level Measured quantity

Before 06/01/1949 0.96b 0.66 Total Pu alpha

01/06/1949–11/30/1952 0.33 None Total Pu alpha

12/01/1952–01/27/1953 0.18 None Total Pu alpha

01/28/1953–03/26/1953 0.15 None Total Pu alpha

03/27/1953–11/06/1953 0.05 None Total Pu alpha

11/07/1953–12/04/1953 0.07 None Total Pu alpha

12/01/1953–04/30/1955 0.057 None Total Pu alpha

05/01/1955–08/30/1955 0.027c None Total Pu alpha

09/01/1955–09/30/1955 0.04c None Total Pu alpha

10/01/1955–09/30/1983 0.05d 0.025d Total Pu alpha

10/01/1983–12/31/1983 0.035 Not established, 

recommend use of 

0.025 consistent 

with previous years 

Each Pu-238, Pu-239

01/02/1984–04/1988 0.02 None Total Pu alpha

05/01/1988–05/1990 0.02 0.01 Total Pu alpha

06/01/1990–11/1991 0.03 0.015 Total Pu alpha

11/01/1991–04/2000 0.02 0.01 Total Pu alpha

05/01/2000–08/2001 0.02 Xb + 2.05 × TPUe Total Pu alpha

09/01/2001–present 0.02 2 × TPUe Total Pu alpha

a. From Wilson [1987] except as otherwise noted.

b. Estimated from data in Table 1 in Bioassay at Hanford [Healy 1948]. 

c. The values are probably closer to a decision level than an MDA. It is recommended that 

dose reconstructors use 0.05 dpm as the MDA [8].

d. During part of this period, results that were less than the detection limit were reported as 

0.025. However, if net activity above background and above 0.025 was detected, the actual 

amount was recorded [9].

e. Xb is mean of blanks and TPU is total propagated uncertainty. The Xb values during that 

time were 0.0000 and 0.000487 dpm for Pu-238 and Pu-239, respectively. However, it is 

recommended that dose reconstructors use 2 × TPU as the actual decision level for both 

periods [10].

The results might have been reported as plutonium or 239Pu but, until October 1983, the result was 

really the total alpha activity from isotopes of plutonium. Before October 1983, the recorded value was 

the total alpha activity from plutonium, so it would have included activity from 238Pu, 239Pu, and 240Pu. 

The recorded results would not have accounted for 241Pu or 241Am present in the urine [11]. Results on 

plutonium urinalysis sheets were recorded in units of disintegrations per minute per sample, but the 

same results were converted to units of microcuries per sample in the database. The units in the 

database should have a unit code of 5, meaning μCi/sample, but if the code is missing or unreadable, 

the units are still recognizable because the exponent is normally -7 or -8. A value of 1.1 × 10-8 was 

recorded for results for which plutonium was not detected (one-half of the nominal 0.05-dpm MDA).
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 26 of 180

This method of recording was used through 1974. In 1975, the units were changed to dpm/sample 

(unit code 1), and 0.025 was recorded for results for which plutonium was not detected [12].

In October 1983, several changes were made. The lanthanum fluoride with TTA method was replaced 

by the use of anion exchange columns, autoradiography was replaced by alpha spectrometry 

analysis, and chemical yield was established for each sample separately by use of a 242Pu tracer. The 

results of 238Pu and 239+240Pu have been reported separately since then. A 2,500-minute count time 

has been used since 1984. A 10,000-minute count time was introduced for special situations in 1996, 

but its use was rare [13].

Starting in the mid-1990s, the fecal procedure was enhanced to ensure improved oxidation of highly 

insoluble plutonium. Added steps included wet ashing with hydrogen peroxide and fusion with 

hydrogen fluoride. This procedure was tested with special high-fired plutonium oxide samples from the 

Idaho National Engineering and Environmental Laboratory and was found to work well [14].

Fecal samples were usually not analyzed in total; they were aliquoted after muffling, dry ashing, and 

wet ashing. Therefore, more than one analysis result for a given sample was possible and occurs

often in the database [15].

The MDAs from 1983 to the present are nominal MDAs based on contractual requirements. In 

general, the laboratory performed slightly better than the contractual MDA, but the true MDA varied 

slightly over time and the contractual MDA was a reliable estimate [16]. Reporting of errors, which 

was the total propagated uncertainty (TPU) including uncertainty in the determination of chemical 

yield, counting efficiency determination, and systematic errors, began in 1981 [17]. The 

implementation of a distinction between an MDA (type I and type II errors) and a decision level (type I 

error) occurred in April 1989 [Lyon et al. 1990]. A fixed value of 0.01 dpm/sample (half the nominal 

MDA) was initially used for the decision level for all results. The decision level was allowed to become 

sample-specific based on the TPU in 2000, and an adjustment was made to the formula in 2001 

[Lynch et al. 2000, 2001].

The MDAs in Table 5-2 apply to routine and priority processing of urine samples. Fecal sampling was 

used for special sampling after potential intakes, and other processing codes (emergency and 

expedite) have been available for special urine and fecal samples. The contractual MDAs for these 

samples are provided in Tables 5-3 and 5-4.

Fecal sampling was normally done in response to suspected intakes, but routine fecal sampling was 

used for some high-risk plutonium workers, mostly operators at the Plutonium-Uranium Extraction 

(PUREX) facility and the Plutonium Finishing Plant (PFP), from 1986 through June 1989. The special 

study showed that, when considered as a group, the mean fecal excretion was statistically 

significantly different from controls. Enhanced air sampling, which was initiated in response to the 

study, showed frequent but intermittent releases of plutonium in the workplaces at levels below the 

detectability of normal air sampling. When modeled as chronic intake, the intakes and doses were low 

(less than 10 mrem committed effective dose equivalent), and were documented in the workers’ 

records [Bihl et al. 1993; Lyon et al. 1988, 1989]. When encountered in the workers’ records, results 

for these fecal samples should be interpreted as associated with chronic intakes, not an acute intake 

occurring many days before the sample dates.

With the exception of a few standards in radiochemistry laboratories and a 238Pu purification 

experiment in the 325 Building C-Cell around 1967 [20], plutonium at Hanford was a mix of 

radionuclides, namely 238Pu, 239Pu, 240Pu, and 241Pu. The activity of 242Pu in plutonium mixtures at 

Hanford was too small to contribute significantly to dose. Hanford plutonium mixtures were 

categorized by their weight percent of 240Pu. When the reactors were operated to produce plutonium 

for weapons, the target mixture was about 6% 240Pu, a mixture referred to as weapons grade.
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Table 5-3. MDAs for nonroutine plutonium fecal analysis 

(dpm/sample) [18]. 

Period Emergencya Expedite Priority

01/1965–10/1983b 0.9–1.5 Not available 0.1–0.15

10/1983–12/1985c 9 Not available 0.2

01/1985–06/1990 9 3 0.2

06/1990–02/1991d 20 4 Not available

02/1991–present 9 3 0.2

a. At times the emergency category was called “rush” and the routine 

category was called “normal.”

b. MDAs varied according to sample size over the range shown; the lower 

value was generally applicable except for very large samples [BMI 1982]. 

MDAs for this period apply to total plutonium alpha.

c. MDAs from this time forward apply to Pu-238 and Pu-239 separately.

d. Emergency and expedited processing of fecal samples was available 

through the PNNL Analytical Chemistry Laboratory. Priority fecal analyses 

were also available through the offsite laboratories but the MDA was not 

established, probably about 0.2 to 0.5 dpm/sample considering the stateof-the-art of those laboratories.

Table 5-4. MDAs for nonroutine plutonium urinalysis 

(dpm/sample) [19]. 

Period Emergencya Expedite

01/1965–10/1983b 0.5–0.7 Not available

10/1983–12/1985c 0.5 Not available

01/1985–06/1990 0.5 0.08

06/1990–02/1991d 2 0.4

02/1991–present 0.5 0.08

a. At times the emergency category was called “rush” and the 

routine category was called “normal.”

b. MDAs varied according to sample size over the range shown; 

the lower value was generally applicable except for very large 

samples [BMI 1982]. MDAs for this period apply to total 

plutonium alpha.

c. MDAs from this time forward apply to Pu-238 and Pu-239 

separately.

d. Emergency and expedited processing of urine samples was 

available through the PNNL Analytical Chemistry Laboratory. 

N Reactor was also operated to produce electrical power for a local public power company. When the 

reactor was operated to produce power, the mixture in the fuel rods when removed from the reactor 

was nominally 12% 240Pu, a mixture referred to as fuel grade. At any given time, individual fuel rods 

would have mixtures that differed from these, as would individual batches of rods starting at the front 

end of the fuel rod dissolution and plutonium extraction processes. However, when refined and 

blended, the target mixture was the weapons-grade mixture [21]. There is evidence at LANL that 

plutonium produced at Hanford in the 1940s had less 240Pu, perhaps closer to 3% [ORAUT 2013], but 

documentation of that at Hanford has not been found. Table 5-5 lists the activities of plutonium 

isotopes and 241Am, which grows in from 241Pu, for 6% 240Pu. Table 5-6 lists the activity ratios for that 

mixture. Table 5-7 lists the activities for 12% 240Pu mixtures, and Table 5-8 lists the ratios [BMI

2003a]. In these tables, “aging” refers to the time since the 241Am was separated from the plutonium 

(and then starts to build in again from decay of 241Pu). 

The values in these tables can help determine the total intake of plutonium and 241Am if there are 

limited data on the composition of the source of the intake. For instance, only in rare large intakes was 

241Pu measured as part of the intake, so the activity of that isotope is almost never available.

Americium-241 at the time of intake was also often not determined directly. Since 1983, 238Pu and 

239+240Pu were measured separately, so the ratio of one to the other can be used to estimate the 
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Table 5-5. Specific activities (Ci/g) of Hanford reference weapons-grade plutonium mixture (6%). 

Mixture designation, years 

of aging (right)a 

and isotope (below)

Fresh, 

0.038 

(2 weeks)

5-yr, 

5

10-yr, 

10

15-yr, 

15

20-yr, 

20

25-yr, 

25

30-yr, 

30

Pu-238 8.56E-03 8.23E-03 7.91E-03 7.60E-03 7.31E-03 7.03E-03 6.75E-03

Pu-239 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02

Pu-240 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02

Pu-241 8.24E-01 6.48E-01 5.09E-01 4.00E-01 3.15E-01 2.48E-01 1.95E-01

Pu-242 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06

Am-241 5.3E-05 5.83E-03 1.04E-02 1.39E-02 1.66E-02 1.87E-02 2.03E-02

Pu-239+240 7.13E-02 7.13E-02 7.13E-02 7.13E-02 7.12E-02 7.12E-02 7.12E-02

Pu-alpha 7.99E-02 7.95E-02 7.92E-02 7.89E-02 7.85E-02 7.83E-02 7.80E-02

Total alpha 7.99E-02 8.53E-02 8.96E-02 9.28E-02 9.52E-02 9.70E-02 9.83E-02

a. Time since separation of Am-241 from the plutonium mix.

Table 5-6. Activity ratios of Hanford reference weapons-grade plutonium mixture (6%). 

Mixture designation, years 

of aging (right)a 

and isotope ratios (below)

Fresh, 

0.038 

(2 weeks)

5-yr, 

5

10-yr, 

10

15-yr, 

15

20-yr, 

20

25-yr, 

25

30-yr, 

30

Pu-239+240:Am-241 1,345 12.2 6.87 5.13 4.28 3.80 3.50

Pu-239+240:Pu-238 8.33 8.67 9.01 9.38 9.74 10.1 10.5

Pu-241:Pu-239+240 11.6 9.09 7.15 5.62 4.42 3.48 2.73

Pu alpha:Pu-239+240 1.12 1.12 1.11 1.11 1.10 1.10 1.10

Pu alpha: Pu-238 9.33 9.66 10.0 10.4 10.7 11.1 11.6

Pu alpha:Am-241 1,507 13.6 7.62 5.68 4.73 4.19 3.84

Pu-241: Pu alpha 10.3 8.15 6.43 5.07 4.01 3.17 2.50

a. Time since separation of the Am-241 from the plutonium mixture.

Table 5-7. Specific activities (Ci/g) of Hanford reference fuel-grade plutonium mixture (12%). 

Mixture designation, years 

of aging (right)a 

and isotope (below)

Fresh, 

0.038 

(2 weeks)

5-yr, 

5

10-yr, 

10

15-yr, 

15

20-yr, 

20

25-yr, 

25

30-yr, 

30

Pu-238 1.71E-02 1.64E-02 1.58E-02 1.52E-02 1.46E-02 1.40E-02 1.35E-02

Pu-239 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.25E-02

Pu-240 2.72E-02 2.72E-02 2.72E-02 2.72E-02 2.72E-02 2.71E-02 2.71E-02

Pu-241 3.09E+00 2.43E+00 1.91E+00 1.50E+00 1.18E+00 9.29E-01 7.30E-01

Pu-242 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06

Am-241 2.0E-04 2.19E-02 3.89E-02 5.22E-02 6.24E-02 7.03E-02 7.63E-02

Pu-239+240 7.98E-02 7.98E-02 7.98E-02 7.97E-02 7.97E-02 7.97E-02 7.97E-02

Pu-alpha 9.69E-02 9.62E-02 9.56E-02 9.49E-02 9.43E-02 9.37E-02 9.32E-02

Total alpha 9.69E-02 1.18E-01 1.35E-01 1.47E-01 1.57E-01 1.64E-01 1.69E-01

a. Time since separation of the Am-241 from the plutonium mix.

Table 5-8. Activity ratios of Hanford reference fuel-grade plutonium mixture (12%). 

Mixture designation, years 

of aging (right)a 

and isotope ratios (below)

Fresh, 

0.038 

(2 weeks)

5-yr, 

5

10-yr, 

10

15-yr, 

15

20-yr, 

20

25-yr, 

25

30-yr, 

30

Pu-239+240:Am-241 399 3.64 2.05 1.53 1.28 1.13 1.04

Pu-239+240:Pu-238 4.67 4.86 5.05 5.24 5.46 5.69 5.90

Pu-241:Pu-239+240 38.7 30.5 24.0 18.8 14.8 11.7 9.16

Pu alpha:Pu-239+240 1.21 1.21 1.20 1.19 1.18 1.18 1.17

Pu alpha: Pu-238 5.67 5.87 6.05 6.24 6.46 6.69 6.90

Pu alpha:Am-241 485 4.39 2.46 1.82 1.51 1.33 1.22

Pu-241: Pu alpha 31.9 25.3 20.0 15.8 12.5 9.91 7.83

a. Time since separation of the Am-241 from the plutonium mixture.
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category of the plutonium mixture and, from the tables, to estimate the activities of 241Pu and 241Am.

Before 1983, the measured quantity was total alpha from plutonium, which means the total of 238Pu 

and 239+240Pu. Therefore, unless 241Am was measured or there is other information about the intake, 

there might be no way to tell from the bioassay how much 241Pu and 241Am were present at intake.

Most plutonium mixtures at Hanford were nominally weapons grade and, if the 239+240Pu-to-238Pu ratio 

from an intake implies weapons grade (e.g., >8), dose reconstructors should use the ratios in 

Table 5-5. However, if no information is available about the nature of the plutonium mixture that is 

associated with an intake, or with less than MDA results used to calculate missed dose, and the 

intake is being calculated from urine sample results, the 12% mixture can be used as a default 

assumption that is favorable to claimants. Because there is less 241Am in the 6% mix than the 

12% mix, use of a 6% mixture is the default starting point for limiting doses based on chest counts. If 

both types of data are available, it is necessary to compare the two.

In relation to the age of the plutonium, until the 5th year of fuel separation (1949), assume fresh 

plutonium. For 1950 through 1954, assume a 5-year-old plutonium mixture. After these times, the 

assumption of 10-year-old plutonium should be made. For intakes since about 1996, a 20-year-old 

fuel-grade mixture could be assumed.

There was at least one project in the 1970s that involved irradiated fuel rods from commercial power 

reactors (Nuclear Waste Vitrification Project). The work took place in the 324 and 325 Buildings and 

the material was stored in the 303C Building in the 300 Area [22]. Commercial fuel rods have a much 

higher degree of burnup, and those at Hanford were characterized by much more 241Pu and nominally 

26% 240Pu. Table 5-9 provides the activities of the commercial fuel used in the Nuclear Waste 

Vitrification Project; Table 5-10 provides the activity ratios. In addition, the PFP sometimes recycled 

plutonium from other DOE sites, the United Kingdom, and commercial reactors. These materials often 

had higher 240Pu and 241 Am content [Hoyt and Teal 2004, pp. 46–50]. Plutonium from the West Valley 

commercial reprocessing site is also stored at Hanford. However, unless the records about the 

specific intakes being investigated have evidence of these unusual mixtures, the default mixtures 

above should be used.

Table 5-9. Specific activities (Ci/g) of Hanford reference commercial power fuel-grade plutonium

mixture [BMI 2003a]. 

Mixture designation, years of 

aging (right)a 

and isotope (below)

Fresh, 

0.038 

(2 weeks)

5-yr, 

5

10-yr, 

10

15-yr, 

15

20-yr, 

20

25-yr, 

25

30-yr, 

30

Pu-238 1.71E-01 1.64E-01 1.58E-01 1.52E-01 1.46E-01 1.40E-01 1.35E-01

Pu-239 3.41E-02 3.41E-02 3.41E-02 3.41E-02 3.41E-02 3.41E-02 3.41E-02

Pu-240 5.90E-02 5.89E-02 5.89E-02 5.89E-02 5.89E-02 5.88E-02 5.88E-02

Pu-241 1.34E+01 1.05E+01 8.28E+00 6.51E+00 5.12E+00 4.03E+00 3.17E+00

Pu-242 1.97E-04 1.97E-04 1.97E-04 1.97E-04 1.97E-04 1.97E-04 1.97E-04

Am-241 8.22E-04 9.49E-02 1.69E-01 2.26E-01 2.79E-01 3.04E-01 3.31E-01

Pu-239+240 9.31E-02 9.31E-02 9.30E-02 9.30E-02 9.29E-02 9.29E-02 9.29E-02

Pu-alpha 2.65E-01 2.58E-01 2.52E-01 2.45E-01 2.39E-01 2.34E-01 2.28E-01

Total alpha 2.65E-01 3.53E-01 4.20E-01 4.71E-01 5.10E-01 5.38E-01 5.59E-01

a. Time since separation of the Am-241 from the plutonium mixture.

Use known ratios if available or default activity ratios to determine the isotopic components of the total 

alpha results before combining the datasets for curve fitting. If some of the plutonium bioassay was 

obtained before October 1983 and some after, the two datasets are not compatible. The results from 

before October 1983 are the sum of activities from 238Pu, 239Pu, and 240Pu; the results from October 

1983 and after are separate for 238Pu and 239+240Pu. 
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Plutonium at Hanford could have existed as absorption types M and S, and as a highly insoluble form, 

type SS [ORAUT 2020]. Dose reconstructors should model the 241Am that is a component of 

plutonium contamination in the lung the same as the plutonium matrix in which it has grown. In other 

words, the americium should be treated as absorption type S if the plutonium is type S.

Table 5-10. Activity ratios of Hanford reference commercial power fuel-grade plutonium mixture 

[BMI 2003a]. 

Mixture designation, years of 

aging (right),a 

and isotopes ratioed (below)

Fresh, 

0.038 

(2 weeks)

5-yr, 

5

10-yr, 

10

15-yr, 

15

20-yr, 

20

25-yr, 

25

30-yr, 

30

Pu-239+240:Am-241 113 0.981 0.551 0.411 0.344 0.305 0.281

Pu-239+240:Pu-238 0.544 0.568 0.589 0.612 0.636 0.664 0.688

Pu-241:Pu-239+240 144 113 89.1 70.0 55.1 43.3 34.1

Pu alpha:Pu-239+240 2.85 2.77 2.71 2.63 2.57 2.52 2.45

Pu alpha: Pu-238 1.55 1.57 1.59 1.61 1.64 1.67 1.69

Pu alpha:Am-241 322 2.72 1.49 1.08 0.857 0.770 0.689

Pu-241: Pu alpha 50.6 40.7 32.9 26.6 21.4 17.2 13.9

a. Time since separation of the Am-241 from the plutonium mixture.

A significant decrease in the number of plutonium bioassays across the Hanford Site occurred starting 

in 1960. This resulted from a change in policy on the frequency of sampling of workers. A 

determination of the risk of intake was made and workers were reassigned to sampling frequencies of 

quarterly, annually, or once every 5 years.

5.2.2 Americium

Americium was usually a trace contaminant in plutonium mixtures, as discussed in Section 5.2.1. 

However, americium was separated from plutonium at the reprocessing plants such as T Plant and 

S Plant (also called the Reduction-Oxidation [REDOX] Plant in the early years and the PUREX Plant 

from 1956) and at the Plutonium Reclamation Facility (a wing in the PFP). There was a major 

americium recovery operation in the Waste Treatment Facility, 242-Z, from 1964 until 1976. Smaller 

scale separation of americium from plutonium occasionally occurred earlier at 231-Z and later at the 

325 Building in the 300 Area. See the latest revision of ORAUT-TKBS-0006-2, Hanford Site – Site 

Description, for details [ORAUT 2010a]. This americium should be treated as americium (rather than 

trace americium atoms bound in a plutonium matrix) [23]. The International Commission on 

Radiological Protection (ICRP) has recommended absorption type M for americium [ICRP 1994a]. 

As presented in Section 5.1.3, an SEC class was designated in which NIOSH found that it was 

infeasible to reconstruct doses for americium separated from plutonium in the 200 Area before 

December 31, 1968, due to inadequate monitoring. Although doses from separated americium cannot 

be reconstructed during this period, doses can be reconstructed for americium that is part of the 

plutonium matrix. Doses for separated americium can be reconstructed if the worker’s individual 

record includes chest counts or americium urinalysis results.

Routine chest counting for 241Am started in 1967, and special chest counts started in 1966 (see 

Section 5.3.2). 

It has not been discovered yet when an americium urinalysis procedure was first developed at 

Hanford. There is no mention of americium excreta analysis in Healy [1948]; no mention in a 1954 

memorandum “Bioassay Annual Report” that lists numbers of urinalyses for plutonium, fission 

products, and uranium [Watters 1954]; and no mention in a compilation of bioassay procedures, 

Bioassay Procedures and Analysis (Old Bioassay Bible) [Bioassay procedures 1961]. 
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The first americium records in the database show 41 urinalyses for 241Am in 1964 on samples that 

were collected from 19 workers. No americium urinalyses were recorded in 1965 or 1966. The 1964 

samples were collected over a short period. The first samples, 12 of them, were collected on 

August 25 or 26. The last sample was collected on October 29. The results range from 1.3 × 10-6 to 

6.4 × 10-6 [24]. No units are given, although the magnitude of the results suggests the units are 

μCi/sample or μCi/L. It is not clear whether the units are per sample or per liter. Hanford used per liter 

from 1967 through July 1969 and, because the volumes can vary from less than a liter to more than a 

liter, no particular choice is technically better than another. Therefore, the assumption that the unit 

was μCi/L is reasonable [25].

One set of four samples for one worker was collected as followup to a contamination event. Among 

the remaining workers, most were sampled on August 25 or 26 and then had at least one additional 

sample a month or two later. Therefore, it is likely that the August 25 and 26 samples were baselines 

to establish the bioassay program at the start of the experiments that were being conducted to 

support a new 241Am separation process at the PFP [26].

No information about the radiochemical analysis method or detection level has been found. Assuming 

the August samples were baselines, a cumulative probability analysis was performed on the results 

that provided a median value of 1.9 × 10-6 μCi/L and a 95th percentile of 2.3 × 10-6 μCi/L. Based on 

this analysis, the MDA for 241Am analyses in 1964 was assumed to be 4.6 × 10-6 μCi/L [27].

The records show no americium bioassay in 1965 or 1966; there are 168 results in 1967. The latter 

are a combination of urine and fecal results, and most of the results seem to be duplicated twice in the 

database. The duplication appears to be a mistake in the database that does not show up in hardcopy listings of these bioassay results. The results are all from special samples for [redacted] workers 

who were involved in the same potential intake accident in [Redacted] on [Redacted]. There are 

internal dosimetry evaluations in the workers’ files that document the accident and bioassay results.

Many of the urine sample results were listed as <5.41 × 10-7 μCi/L, so it is assumed that this was the 

MDA for the analysis. This is the same number that was handwritten in the margin of an unpublished 

article, Evaluation of Internal Depositions of Americium Using Bioassay Samples [Henle 1968]. The 

handwritten note says, “5.41 × 10-7 μCi/L detection level.”

A memorandum to file from J. J. Jech [1969], Senior Development Engineer in the Personnel 

Dosimetry Services, states that according to a telephone conversation with M. M. Lardy at UST, the 

new detection limit for 241Am was 2.0 dpm/sample as of July 10, 1969. Lardy described the procedure 

as dibutyl-N,N-diethylcarbamylphosphonate (DDCP) extraction to a planchet and gross-alpha 

counting [ORAUT 2003]. According to a letter from Lardy to H. Larson, manager of Personnel 

Dosimetry Services, a reduction in the detection limit was implemented in 1974 [Lardy 1974]. The new 

limit for 241Am in urine was given as 0.1 pCi/sample at the 90% confidence limit. This limit was still 

listed in a statement of work with UST in 1979 and again in 1982, although it was stated as 

0.2 dpm/sample. In the laboratory statement of work for a new contract starting October 1983 

[Simpson 1983], the detection level was listed as 0.04 dpm/sample. This improvement was achieved 

by use of an alpha/gamma coincidence counter.

Until October 1983, the gross alpha count could have included 242Cm or 244Cm if any was associated 

with the intake. Assuming the results are 241Am is favorable to claimants. However, sometime 

between October 1983 and October 1985, both the chemistry procedure and the counting technique 

were changed. The chemistry method was similar to that in the Health and Safety Laboratory 300 

manual [Volchok and de Planque 1982] and commonly referred to as the “RICH-RC-50-80” method 

[ORAUT 2003]. This method involved sequential precipitation with calcium oxalate and iron hydroxide, 

removal of plutonium using anion exchange, loading on another column with nitric acid and methanol, 

and elution of the americium with HCl and methanol. Electrodeposition and counting by alpha 

spectrometry were also implemented at this time [ORAUT 2003]. The MDA in the 1985 statement of 
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work was listed as 0.02 dpm/sample (consistent with the change to alpha spectrometry) [DOE 1985], 

and it has stayed there to the present. At present, Eichrom transuranic column exchange is used for 

the separation of the americium for urine, but the MDA is the same [ORAUT 2003]. Table 5-11 

summarizes what has been uncovered about 241Am MDAs for routine urinalysis.

Table 5-11. Routinea 241Am urinalysis detection levels [28]. 

Period MDA Decision level

01/1964

10 dpm/L

(4.6E-6 μCi/L)

Anything detected

01/1967–06/1969

1.2 dpm/L

(5.4E-7 μCi/Lb)

Anything detected

07/1969–02/1974 2.0 dpm/sample Anything detected

03/1974–10/1983 0.2 dpm/sample Anything detected

10/1983–09/1985 0.04 dpm/sample Anything detected

10/1985–05/1988 0.02 dpm/sample Anything detected

05/1988–06/1990 0.02 dpm/sample 0.01 dpm/sample

07/1990–10/1991 0.03 dpm/sample 0.015 dpm/sample

11/1991–04/2000 0.02 dpm/sample 0.01 dpm/sample

05/2000–08/2001 0.02 dpm/sample (Xb + 2.05 × TPUc) dpm/sample

09/2001–present 0.02 dpm/sample (2 × TPU) dpm/sample

a. At times the routine category was called “normal.”

b. Not known if MDA or decision level.

c. Xb is the mean of the blanks and TPU is total propagated uncertainty. The Xb value 

used during that time was 0.00276 dpm for Am-241. However, it is recommended that 

dose reconstructors use 2 × TPU as the actual decision level for both periods.

The MDAs in Table 5-11 apply to routine and priority processing of urine samples. Fecal sampling 

was used for special sampling after potential intakes, and other processing codes (emergency and 

expedite) have been available for special urine and fecal samples. The contractual MDAs for these 

samples are provided in Tables 5-12 and 5-13. These analyses could have been used because of 

suspected intakes of pure 241Am (such as the explosion of an americium exchange column at the PFP 

in 1976) or to determine the activity of 241Am in a plutonium mixture. There is evidence of a few 

intakes of pure 241Am before 1969 that involved unusual circumstances such as using a supposedly 

sealed source that had ruptured. These intakes were analyzed by urinalysis, so a procedure obviously 

existed at that time, although it is not part of the contract with UST [29]. On rare occasions for a 

serious intake, samples were analyzed for 241Am using an LEPD before any chemistry. This technique 

came into existence in 1986 or 1987. Its detection level was about 5 dpm/sample [DOE 1987]. In 

general, the LEPD result was used just as a rapid indicator, and a more accurate result was obtained 

by wet chemistry and alpha spectrometry days later [30].
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Table 5-12. MDAs for nonroutine 241Am fecal analysis (dpm/sample) [31].

Period Emergencya Expedite Priority

01/1967–02/1974 (b) Not available (b)

02/1974–12/1981 (c) Not available 4

01/1982–09/1983 3.6–12 (3.6 most probable)d Not available 1.2–5.0 (1.2 most probable)d

10/1983–09/1985 200 Not available 0.16

10/1985–06/1989 20 6 0.1

07/1989–10/1991e 20 4 Not available

11/1991–present 20 6 0.1

a. At times the emergency category was called “rush.” 

b. Probably available but MDAs not found.

c. Emergency analyses were available on request, but the statement of work (based on 1978 statement of work 

[DOE 1979]) did not specify the MDAs. It implied that an MDA about 10 times the routine (or priority for fecal) 

MDA was expected.

d. Varied according to sample size over the range shown; the lower value was generally applicable except for 

very large samples [BMI 1982]. 

e. Emergency and expedited processing of urine and fecal samples was available through the PNNL Analytical 

Chemistry Laboratory. Priority fecal analyses were also available through the offsite laboratories, but the MDA 

was not established; it was probably about 0.2 to 0.5 dpm/sample considering the state of the art of those 

laboratories.

Table 5-13. MDAs for nonroutine 241Am fecal urinalysis 

(dpm/sample) [32].

Period Emergencya Expedite

01/1967–02/1974 (b) Not available

02/1974–12/1981 (c) Not available

01/1982–09/1983 0.7–1.0

(0.7 most probable)d

Not available

10/1983–09/1985 1.0 Not available

10/1985–06/1989 1 0.08

07/1989–10/1991e 2 0.4

11/1991–present 1 0.08

a. At times the emergency category was called “rush.”

b. Probably available but MDAs not found.

c. Emergency analyses were available on request, but the 

statement of work (based on 1978 statement of work [DOE 

1979]) did not specify the MDAs. It implied that an MDA about 

10 times the routine (or priority for fecal) MDA was expected.

d. Varied according to sample size over the range shown; the 

lower value was generally applicable except for very large 

samples [BMI 1982]. 

e. Emergency and expedited processing of urine and fecal 

samples was available through the PNL Analytical Chemistry 

Laboratory. Priority fecal analyses were also available through 

the offsite laboratories, but the MDA was not established; it 

was probably about 0.2 to 0.5 dpm/sample considering the 

state of the art of those laboratories.

5.2.3 Curium

The curium isotopes of concern were 242Cm and 244Cm. Most sources of curium at Hanford were 

minor calibration sources or minor constituents in an actinide mixture. However, flowsheet 

development for major extraction of 244Cm occurred in the 325 Building in 1966 and 1967 and led to 

major extraction and purification of curium from the Shippingport waste in 201-C and 325 Buildings in 

1967 and 1968. Additional work with curium occurred in the 308 Building in smaller quantities in 1969 

through 1973. See the latest revision of the Hanford Site Description for details [ORAUT 2010a]. 
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The curium and americium procedure was the same, so the results would have been reported as 

curium only if so requested through the bioassay request system until alpha spectrometry was 

initiated in October 1983. After 1985, the chemistry was the same as americium, but 241Am, 242Cm, 

and 244Cm were reported separately on request [DOE 1987]. However, the MDAs were not always 

identical with 241Am. Routine urinalysis MDAs for curium are provided in Table 5-14. Nonroutine 

excreta analyses are provided in Tables 5-15 and 5-16 for fecal and urine samples, respectively.

Table 5-14. Routinea curium urinalysis detection levels (dpm/sample) [33]. 

Period MDA Decision level

07/1969–12/1981 Not specifically mentioned Not specifically mentioned 

01/1982–09/1913 Listed for emergency processing only Listed for emergency processing only

10/1983–04/1988 0.02 Anything detected

05/1988–06/1990 0.02 0.01

06/1990–10/1991 0.03 0.015

11/1991–04/2000 0.02 0.01

05/2000–08/2001 0.02 Xb + 2.05 × TPUb

09/2001–present 0.02 2 × TPU

a. At times the routine category was called “normal.”

b. Xb is the mean of the blanks and TPU is total propagated uncertainty. The Xb values used during that time 

were 0.00206 and 0.0000 dpm for Cm-242 and Cm-244, respectively. However, it is recommended that dose 

reconstructors use 2 × TPU as the actual decision level for both periods [34].

Table 5-15. MDAs for nonroutine curium fecal analysis (dpm/sample) [35]. 

Period Emergencya Expedite Priority

Before 1982 (b) Not applicable (b)

01/1982–09/1983 10c Not applicable Not applicable

10/1983–09/1985 240 Not applicable 0.8

10/1985–06/1989 240 70 0.8

07/1989–10/1991d Not applicable Not applicable Not applicable

11/1991–present 240 70 0.8

a. At times the emergency category was called “rush.”

b. Probably available but MDAs not found.

c. Total alpha; would have included any americium present.

d. Emergency and expedited processing of urine and fecal samples was available through the 

PNNL Analytical Chemistry Laboratory. Priority fecal analyses were also available through offsite 

laboratories, but the MDA was not established; it was probably about 0.2-0.5 dpm/sample 

considering the state of the art of those laboratories.

Table 5-16. MDAs for nonroutine curium urinalysis (dpm/sample) 

[36]. 

Period Emergencya Expedite

Before 1982 (b) Not applicable

01/1982–09/1983 0.5–1.0 (0.5 most probable)c,d Not applicable

10/1983–09/1985 10 Not applicable

10/1985–06/1989 1 1.2

07/1989–10/1991e Not applicable Not applicable

11/1991–present 1 1.2

a. At times the emergency category was called “rush.”

b. Probably available but MDAs not found.

c. Total alpha; would have included any americium present.

d. Varied according to sample size over the range shown; the lower value was 

generally applicable except for very large samples [BMI 1982]. 

e. Emergency and expedited processing of urine and fecal samples was 

available through the PNL Analytical Chemistry Laboratory. Priority fecal 

analyses were also available through offsite laboratories, but the MDA was 

not established; it was probably about 0.2-0.5 dpm/sample considering the 

state of the art of those laboratories.
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5.2.4 Tritium

Tritium was referred to as P-10 in the 1950s [Parker 1950a]. The main source of tritium in the 1950s 

was 108-B, also called the P-10 Plant, in which tritium was extracted from irradiated lithium-aluminum 

targets intermittently between August 1949 and 1955. A 1967 report states [McConnon 1967]:

Battelle-Northwest and its predecessor at Hanford, the General Electric Company, have 

been involved in activities with tritium since about 1950, initially as a manufactured 

product for weapons applications and later as a by-product of heavy water reactor 

operations. Our most recent experience is from operation of the Plutonium Recycle Test 

Reactor (PRTR).

The PRTR started up in 1960 and was shut down in 1969. All of the major reactors had some 

potential for tritium exposure, and tritium was used in tracer quantities in biological research through 

most of Hanford’s history. There was also some work on a tritium target program in the 1990s in the 

300 Area and tritium light sources in the 1980s (involving just a few people). Tritium in the 

groundwater was a source of ingestion for 400 Area workers, principally the Fast Flux Test Facility 

(FFTF), from about 1978 to 2003 [ORAUT 2005a]. PNNL has participated in the National Nuclear 

Security Administration Tritium Readiness Program in recent years, producing and testing tritiumproducing burnable absorber rods in the 324 and 325 Buildings [Lanning 2007]. 

The history of tritium urinalysis at Hanford is not well documented. Tritium urinalysis was not 

mentioned at all in Wilson [1987]. The earliest report on tritium urinalysis at Hanford was made in

1949 by J. Healy, the leading internal dosimetrist at Hanford for many years [Healy 1949]. That 

procedure was based on “production of acetylene from the active water, with subsequent 

measurement of the ionization caused by the tritium beta particle” [Healy 1949]. No detection level 

was mentioned in that letter, but one was mentioned in a memorandum that referred to the acetylene 

method for urinalysis; it provides a sensitivity of about 1.2 μCi/L in water [Parker 1950a]. However, 

that method apparently did not work well because a 1951 letter stated that, “Your problem on the 

determination of tritium in the urine samples is one that we have been working on for the last two 

years, and have finally obtained what appears to be a decent method for routine use” [Healy 1951]. 

The copy of the letter is of such poor quality that the method is hard to follow, but it definitely was not 

liquid scintillation counting. Tritium urinalysis for the workers at the Tritium Extraction Facility (108B) 

was conducted at 108B until 1955 when the counter was moved to the bioassay laboratory in the 

329 Building [Watters 1955, p. 72]. A 1961 report, The Estimation of Whole Body Dose from Tritium 

by Urine Analysis, indicated that liquid scintillation was used by that time, but again no detection level 

was given [Beasley and Rouse 1961, p. 8]. Liquid scintillation counting was implemented for tritium 

bioassay at the Savannah River Site in 1958, and it is reasonable to expect that Hanford did so at 

about the same time. In an interview, a former employee stated that liquid scintillation counting of a 

1-mL aliquot of raw urine had been used since UST was awarded the bioassay contract in 1965

[ORAUT 2003]. 

Tritium intakes were accounted for as part of external dose until about 1987, when they were entered 

in the dose database as an internal dose [37]. Section 5.6 contains a more detailed discussion and 

guidelines for assigning intakes. Tritium exposure was assumed to be chronic during the exposure 

period, unless a very large acute intake was known to occur.

Very little data on MDAs has been discovered. A sensitivity of 1.2 μCi/L was reported by Parker in 

1950, but what constituted sensitivity was not mentioned [Parker 1950a, p. 4]. McConnon [1964] 

states that a tritium bioassay result exceeding 5 μCi/L will be reported to the Radiation Monitoring 

Office the day after the samples are picked up, which indicates a level of concern probably well above 

the MDA. One P-10 Personnel Sample Analysis card, with entries in 1952, shows several values 

below 5 μCi/L and the smallest as 2.5 μCi/L. None of the values is listed as a less-than value. The 
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1965 statement of work with UST [AEC 1964] shows an MDA of 1 μCi/L, which is consistent with the 

MDA at Savannah River throughout the 1950s. Table 5-17 lists the most accurate compilation of 

MDAs for routine tritium urinalysis to date. From 1978 to the present, the MDAs were obtained from 

statements of work with the bioassay laboratory.

Table 5-17. Routine tritium urinalysis detection 

levels [39]. 

Period MDA

01/1949–12/1960a ≈5 μCi/L

01/1961–12/1981b 1 μCi/L

01/1982–10/1991 0.0045 μCi/L 

(10 dpm/mL)

11/1991–present 0.009 μCi/L 

(20 dpm/mL)

a. Dates and MDAs are best guesses. The change in 1961 

was based on earliest reference to liquid scintillation 

counting.

b. The reporting level continued to be 5 μCi/L at least 

through August 1964 [GE 1964, p. 151]. 

Tritium was present in an organic form in the 108-B facility in the pump oil, which became 

contaminated over time. According to a retired radiation monitor at the facility, the contamination in 

pump oil was orders of magnitude below general tritiated gas (HT) and tritiated water vapor (HTO) 

levels in the facility [38]. The tritium in the lithium-aluminum rods was absorbed in the aluminum as a 

getter, so it would have been present as a metal tritide, albeit a weak one. An article written at the 

time of operation of 108-B stated that “aluminum is one of the metals least likely to absorb tritium” 

[Kornberg 1950]. Therefore, the tritium was easy to drive off as part of the process. Workers did not 

handle the aluminum directly and, after the tritium was driven off, the aluminum was placed in cans, 

sealed, and buried. This work was done in gloveboxes. There was also contaminated mercury in the 

vacuum lines to which maintenance workers had access about twice a year; special precautions were 

taken whenever workers had to be exposed to mercury because of the high contamination [40].

Metal tritides were potentially present as part of the Tritium Target Program work starting in 1988; the 

mostly likely metal was zirconium. The irradiated targets were examined and the tritium was driven off 

in hot cells. Samples were taken for analysis, but where the analyses were performed has not been 

discovered yet.

5.2.5 Uranium

As presented in Section 5.1.3, a class was added to the SEC for all employees from July 1, 1972, 

through December 31, 1983. NIOSH found that it was infeasible to reconstruct doses for HEU and for 

233U from July 1, 1972, through December 31, 1983, due to inadequate monitoring.

Uranium exposure at Hanford involved principally three physical forms: depleted uranium (DU), 

natural uranium (NU), and slightly enriched (also called recycled) uranium (RU). Uranium-233 was 

also isolated from irradiated thorium at PUREX from 1966 to approximately 1971. Small numbers of 

researchers might have experimented with more-enriched uranium at different times (e.g., metallurgy 

on commercial-grade fuel), but such exposure would have been to small groups for limited periods. 

For instance, there were radiation work permits for a variety of enriched uranium materials in the early 

1970s [Jech 1970a]. Tables 5-18 and 5-19 provide the default uranium mixtures (RU and commercial 

fuel, respectively) that have been used at the Hanford Site since the 1980s [BMI 2003a]. However, 

different batches of uranium can have different isotopic mixtures. Therefore, the default mixtures in 

the Integrated Modules for Bioassay Analysis (IMBA) program and the Project dose reconstruction 

tools are acceptable. In general, personnel working in the production facilities (e.g., fuel fabrication, 
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the reactors, fuel dissolution and plutonium processing, and waste management) were exposed to NU 

during operation of the early reactors (through 1951) and to RU starting in February 1952 at the UO3

Plant and July 1952 at the 300 Area Fabrication plants [DOE 2000]. RU has impurities that build up 

and track with the uranium over time. Recommended impurity levels based on tolerance specifications 

or higher than normal concentrations in some drums are provided in Table 5-20 [DOE 2000]. 

Plutonium-239 can be assumed for the plutonium alpha impurity, and 232Th can be assumed for the 

thorium [41].

Table 5-18. Radiological characteristics of Hanford RU.

a,b

Radionuclide

Weight 

percentage

Specific constituent 

activity in mixture

(μCi/g, nCi/mg, or pCi/μg)c

Specific constituent 

activity in mixture 

(dpm/μg)c

Constituent fraction 

of total uranium 

activity in mixture

U-234 0.0082 0.5125 1.1378 0.5632

U-235 0.9700 0.0210 0.0465 0.0230

U-236 0.0680 0.0440 0.0977 0.0484

U-238 98.9500 0.3325 0.7381 0.3654

Totals 99.9962 0.9099 2.0200 1.0000

a. RU fuel data from BMI [2003a]. 

b. RU data based on average of data presented by Sula et al. [1991]. 

c. Can be used to represent specific alpha activity in the mixture as well.

Table 5-19. Radiological characteristics of Hanford commercial fuel.

a,b

Radionuclide

Weight 

percentage

Specific constituent 

activity in mixture

(μCi/g, nCi/mg, or pCi/μg)c

Specific constituent 

activity in mixture 

(dpm/μg)c

Constituent fraction 

of total uranium 

activity in mixture

U-234 0.0300 1.8750 4.1625 0.8279

U-235 2.9600 0.0639 0.1419 0.0282

U-236 Negligible Negligible Negligible Negligible

U-238 97.0100 0.3260 0.7236 0.1439

Totals 100 2.2649 5.0281 1.0000

a. Commercial fuel data from Rich et al. [1988]. 

b. RU data based on average of data presented by Sula et al. [1991]. 

c. Can be used to represent specific alpha activity in the mixture as well.

Table 5-20. Impurities in RU at Hanford. 

Constituent Maximum alloweda Observed rangeb Recommended levelc

Pu 10 ppb U <1–2 ppb U 0.8 nCi Pu-alpha/g U

Np Not established 0.04–0.16 ppm U 0.4 nCi Np-237/g U

Th 750 ppm U 8–10 ppm U 5 pCi Th-232/g U

Tc-99 Not established 3–4 ppm U 0.2 μCi Tc-99/g U

Ru-103/106 <20 μCi/lb U <6 μCi/lb U 40 nCi Ru-106/g U

Zr/Nb-95 <10 μCi/lb U <4 μCi/lb U 20 nCi ZrNb-95/g Ud

Other gamma emitters <2 μCi/lb U 0.09–0.75 μCi/lb U Negligible 

a. From UO3 Plant operating specifications [Sula et al. 1991]. 

b. From analysis of uranium lots 88-1, 88-2, and 88-3 that were processed in 1988, and lots 93-01, 

93-02, 93-03, 93-04, and 93-05, processed in 1993.

c. The recommended levels are expected to result in a slight overestimate of dose compared to levels 

actually observed as presented in the DOE investigation of RU at Hanford [DOE 2000]. The 

plutonium reference level was based on the 10-ppb specification, which was reached or exceeded in 

a few drums throughout Hanford history [DOE 2000]. 

d. Interpret as 10 nCi each of the two radionuclides [42].

Uranium compounds at Hanford ranged from very soluble uranyl nitrate and soluble UO3 to relatively 

insoluble UO2 and U3O8. Dissolution tests in simulated lung fluid were conducted on samples from the 

major uranium handling facilities. Results are shown in Table 5-21. Because the relationship between 

the old lung fluid studies and the ICRP Publication 66 absorption types [ICRP 1994a] is not 

established, Table 5-21 also shows approximate Publication 66 absorption types for intakes from the 
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listed facilities. The information from the lung fluid studies indicates that all material types are present 

on the site. Unless there is claim-specific information about the solubility type and due to the 

uncertainty of the exact location of the worker at all times, dose reconstructors should use the 

solubility type that is favorable to the claimant. These absorption type assumptions should be applied 

to the impurities as well.

Table 5-21. Inhalation class for Hanford uranium compounds. 

ICRP 30a inhalation 

class from lung fluid 

studies Compound and location

Approximate 

ICRP 66b lung 

absorption type 

80% D

20% W

Hanford UO3 Plant smear sample dissolution study in 1984c (UO3

powder)

(d)

10% D

90% Y

Hanford 303-M Building air sample dissolution studye (300 Area 

Uranium Fuel Production Facilities)

(d)

29% D

71% Y

Hanford 333 Building air sample dissolution studye (300 Area 

Uranium Fuel Production Facilities)

(d)

20% D

80% Y

Hanford 306-W Building Machine Shop air sample dissolution 

studye

(d)

Not applicable Uranyl nitrate at PUREX or UO3 Plant F

Not applicable UCl4 Mf

Not applicable U carbonate (assumed form after discharge to the soil) Ff,g

a. ICRP [1981]. 

b. ICRP [1994a]. 

c. Sula et al. [1991].

d. Because the conversions from the solubility studies to the ICRP [1994a] absorption types are not exact, the dose 

reconstructor can use the same percentages for D to F, W to M, etc., or can just use the predominant form to maximize 

dose to the organ of concern; for instance, the 303-M Building uranium might be considered 10% F, 90% S, or all 

type S. 

e. Fisher [1986]. 

f. Long [1993]. 

g. Cooke and Holt [1974] [43].

A note about sampling of UO3 Plant workers: because chemical toxicity was the principal concern for 

uranium exposures at the UO3 Plant, one sampling scheme was to obtain both a Friday evening 

sample and Monday morning sample. This sampling scheme was in use as early as the 1950s and 

possibly earlier. This scheme was changed to Monday morning only sampling in about the early 

1980s. The changeover should be clear in the records.

The Friday and Monday sampling scheme was also used in 1962 and 1963 for 313 and 314 Building 

workers. Dose reconstructors need to be aware of Monday morning only samples and make 

appropriate adjustments so the intake will not be underestimated. For Friday and Monday sampling 

schemes, the Monday samples can be ignored and the Friday samples can be used.

Wilson [1987] states that the uranium urinalysis program before 1948 was not reliable. The 

fluorometric method, which fused uranium from raw urine with sodium fluoride and measured the 

fluorescence when the compound was exposed to ultraviolet light, was implemented sometime during 

the first half of 1948 [Healy 1948; Wilson 1987]. This method was used for elemental uranium 

analyses, with refinements over the years including some upfront chemistry on the raw urine, until 

about 1991 when it was replaced by kinetic phosphorescence analysis (KPA) [ORAUT 2003]. (Note: a 

former employee said “about 1990,” but other evidence indicates late 1991.) Lardy [1970] describes 

two procedures: (1) wet ashing with nitric acid and hydrogen peroxide, then acidification and counting 

of a 100-μL aliquot with a detection level of 0.5 μg/L, and (2) extraction (after wet ashing) with methyl 

isobutyl ketone and ammonium hydroxide. The detection limit for the latter was listed as 0.05 μg/L, but 

the recoveries were about the same for both methods, so the latter must have used a 10-times larger 

aliquot. Based on requirements in later statements of work, it is assumed that the first method was 

used for routine analyses. Lardy [1970] also describes a third method – a radiometric procedure that 
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uses the same separation chemistry as the second procedure – but the sample “is measured by a gas 

flow proportional counter or a ZnS (Ag) scintillation counter.” The detection limit was given as

0.5 dpm/sample. A 1989 description of the chemistry was wet ashing with HCl and extraction with 

hexone [Carbaugh et al. 1989]. A 100-mL aliquot was used, but the results were reported as per total 

sample. The chemistry for the KPA involved a 50-mL aliquot that was wet-ashed with acid, passed 

through an ion exchange column, then eluted with weak acid. Results were reported as per total 

sample.

When alpha spectrometry was introduced in 1983, two uranium urinalyses procedures were offered:

the elemental procedure above and the alpha spectrometric procedure to provide isotopic results. In 

general, the elemental procedure was used for workers who were exposed to natural or slightly 

enriched forms of uranium, and the isotopic procedure was used for depleted or more than slightly 

enriched forms of uranium. In general, personnel who worked in the production facilities were 

monitored by the elemental analysis, whereas PNNL workers were monitored by the isotopic analysis 

because of the wide scope of research projects that occurred over the years [44].

Alpha spectrometry cannot differentiate between 233U and 234U. Before 1994, the results for this region 

of the alpha spectrum were reported as 233U; they have been reported as 234U from 1994 to the 

present unless it was specifically determined that the worker was exposed to 233U. Work with 233U was 

rare after the early 1970s, long before alpha spectrometry came into use for bioassay. Therefore, 

unless 233U is specifically mentioned in an intake investigation report, assume that 233U results since 

1983 are actually 234U [45].

Table 5-22 summarizes the routine urinalysis detection levels for elemental methods. Table 5-23 

summarizes detection levels for isotopic methods. Tables 5-24 and 5-25 summarize nonroutine 

detection levels for fecal analysis and urinalysis, respectively. Although a detection level is presented 

for 1948, from December 1943 through 1948, intakes for uranium should be assigned in accordance 

with Section 5.6.2, which assigns intakes based on the Process/Job Title listed as Machining/Operator 

as presented in Battelle-TBD-6000, Site Profiles for Atomic Weapons Employers That Worked 

Uranium Metals [BT 2011]. 

Starting about 1995, mass spectrometry has been used as an investigational tool to discriminate 

between natural background uranium and RU through measurement of 236U. The presence of 236U 

confirms an occupational intake of RU; the detection limit for 236U is such that urinary excretion of 

uranium greater than 0.2 μg/L (see discussion of natural background excretion below) from an intake 

of RU should have a detectable amount of 236U [BMI 2003a; MacLellan et al. 1995]. 

NU from nonoccupational intakes (primarily food and water) is excreted in urine at levels above the 

analytical MDAs for either the elemental uranium analysis or the alpha spectrometry analysis. The 

234U-to-238U ratio can be used to distinguish DU from NU but, considering uncertainties in analytical 

results, that ratio cannot be used to distinguish RU. Three studies were conducted in 1985, 1990, and 

1995 to establish the range of natural background excretion in unexposed persons who lived near the 

Hanford Site. The third study purposely looked for possible geographic and seasonal differences in 

the background. All studies found natural excretion to have a lognormal distribution. Although the 

50th percentiles and slopes of the excretion curves were different in the studies, each study found 

0.2 μg/d to be between about the 99th and 99.9th percentiles. The 1995 study had one result that 

greatly exceeded the 0.2 μg/d value [BMI 2003a]. 
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Table 5-22. Routinea uranium urinalysis detection levels, elemental. 

Period MDA Decision level

Before 1948 Not specifically mentioned Not specifically mentioned

01/1948–12/1949 10 μg/L Anything detected

01/1950–12/1962b 4 μg/L(reporting level) (c)

01/1963–08/1964 0.14 μg/L(reporting level)d,e Not known

09/1964–12/1973 4 μg/Ld Not known

01/1974–12/1981 0.4 μg/L Same as MDA

01/1982–09/1983 0.05–0.25 μg/L

(0.1 μg/L most probable)f

Same as MDA

10/1983–12/1983 0.03 μg/sample 0.5 μg/sampleg

01/1984–08/1985 0.03 μg/sample Not known

09/1985–06/1990 0.03/0.5 μg/sampleh 0.2 μg/samplei

06/1990–10/1991 0.2/0.5 μg/sampleh 0.2 μg/samplei

11/1991–present 0.06/0.5 μg/sampleh 0.2 μg/sampleg

a. At times the routine category was called “normal.”

b. Wilson [1959]. Lardy [1970] also indicates that the MDA from at least 1970 on was 

0.5 μg/L, but the database shows that the use of 4 μg/L continued as a reporting level.

c. In 1958 to 1962 values were reported well below the 4-μg/L value, but a statement of the 

actual reporting level was not found.

d. GE [1963a, p. 183]; GE [1964, p. 152]. 

e. The detection level of 4 μg/L was established in the first contract with UST [AEC 1964], so 

it was applied starting in 1965 and that is probably when the level changed from 0.14 μg/L. 

However, it is favorable to claimants to extend the 4-μg/L value to September 1964.

f. MDAs were based on sample size, but 0.1 μg/L was applied to most sample sizes [BMI 

1982]. The database shows a change to reporting in μg/sample on July 2, 1982.

g. This is actually an action level. Values below this were recorded but not followed up as 

occupational intakes.

h. The larger value is the MDA for a special (rapid) analysis for UO3 Plant workers based on 

potential chemical toxicity. The need for this special analysis ceased in 1994 after the last 

processing in the UO3 Plant [46].

i. Based on upper level for natural background excretion. See text for discussion.
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Table 5-23. Routinea uranium urinalysis detection levels, isotopic.

Period MDA Decision level

Before 1948 Not available Not available

01/1948–12/1949 Not available Not available

01/1950–12/1962b Not available Not available

01/1963–08/1964 Not available Not available

09/1964–12/1973 Not available Not available

01/1974–12/1981 Not available Not available

01/1982–09/1983 Not available Not available

10/1983–12/1983 0.035 dpm/sample Same as MDA

01/1984–08/1985 0.02 dpm/sample Same as MDA

09/1985–06/1990 0.02 dpm/sample Same as MDA

06/1990–10/1991 0.03 dpm/sample 0.15/0.015 dpm/samplec

11/1991–present 0.02 dpm/sample 0.15/0.010 dpm/samplec

a. At times the routine category was called “normal.”

b. Wilson [1959]. Lardy [1970] also indicates that the MDA from at least 1970 on was 0.5 μg/L, but 

the database shows that the use of 4 μg/L continued as a reporting level.

c. These are actually action levels. The first value applied to U-234 and U-238; second value applied 

to U-235 based on natural background in urine. In 2002, the U-235 decision level was lowered to 

0.007 dpm [BMI 2003b]. 

Table 5-24. MDAs for nonroutine uranium fecal analysis [48]. 

Period

Analysis 

type Emergencya Expedite Priority

1980b Elemental Not available Not available Not available

10/1983–09/1985 Elemental 8 μg/sample Not available 0.3 μg/sample

10/1983–09/1985 Isotopicc 12 dpm/sample Not available 0.3 dpm/sample

10/1985–present Elemental 8 μg/sample 5 μg/sample 0.3 μg/sample

10/1985–present Isotopic 12 dpm/sample 4 dpm/sample 0.3 dpm/sample

06/1990–10/1991d Elemental 20 μg/sample 4 μg/sample Not available

06/1990–10/1991d Isotopic 20 dpm/sample 4 dpm/sample Not available

11/1991–present Elemental 8 μg/sample 5 μg/sample 0.3 μg/sample

11/1991–present Isotopic 12 dpm/sample 4 dpm/sample 0.3 dpm/sample

a. At times the emergency category was called “rush.”

b. Earliest mention found in a contract with the laboratory. The ability to perform analyses on fecal samples was also 

mentioned, but an MDA was not specified.

c. U-234, U-235, and U-238.

d. Emergency and expedited processing of urine and fecal samples was available through PNNL’s Analytical Chemistry 

Laboratory. Priority fecal analyses were also available through the offsite laboratories, but the MDA was not established. 

It was probably about 0.2 to 0.5 dpm/sample considering the state of the art of those laboratories.
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Table 5-25. MDAs for nonroutine uranium urinalysis [49]. 

Period Analysis type Emergencya Expedite

1980b Elemental 10 μg/L Not available

10/1983–09/1985 Elemental 7 μg/sample Not available

10/1983–09/1985 Isotopicc 1 dpm/sample Not available

10/1985–present Elemental 7 μg/sample 0.5 μg/sample

10/1985–present Isotopic 1 dpm/sample 0.12 dpm/sample

06/1990–10/1991d Elemental 20 μg/sample 4 μg/sample

06/1990–10/1991d Isotopic 2 dpm/sample 0.4 dpm/sample

11/1991–present Elemental 7 μg/sample 0.5 μg/sample

11/1991–present Isotopic 1 dpm/sample 0.12 dpm/sample

a. At times the emergency category was called “rush.”

b. Earliest mention found in a contract with the laboratory. The ability to perform 

analyses on fecal samples was also mentioned, but an MDA was not specified.

c. U-234, U-235, and U-238.

d. Emergency and expedited processing of urine and fecal samples was available 

through PNNL’s Analytical Chemistry Laboratory. Priority fecal analyses were also 

available through the offsite laboratories, but the MDA was not established. It was 

probably about 0.2 to 0.5 dpm/sample considering the state of the art of those 

laboratories.

Therefore, 0.2 μg/d was established in 1985 and continues to be used at present as the 

environmental decision level for exposures to NU or RU. Only urinary excretion values greater than 

0.2 μg/d, which converts to 0.15 dpm/d for 234U and 238U and 0.007 dpm/d for 235U, are considered 

indicative of a potential occupational source. Nevertheless, the one result in the 1995 study and many 

worker-specific investigations of urinary results exceeding 0.2 μg/d have shown that results well 

above the environmental screening level do occur from natural sources. Some of these were shown to 

be due to a specific home water well; others occurred from workers on city water from wells (but 

apparently not all wells).

Although natural background levels of uranium have been and are present, dose reconstructors 

should consider any detected uranium to be due to occupational exposure and should not make 

corrections for natural background levels.

Background excretion of uranium in feces probably varies over an even larger range than urinary 

excretion, but a definitive study for the Hanford area has not been conducted. Fecal samples were 

rarely obtained for potential uranium intakes; when they were, the investigation report should discuss 

how the results were interpreted.

As presented in Section 5.1.3, a class was added to the SEC for all employees from July 1, 1972, 

through December 31, 1983. NIOSH found that it was infeasible to reconstruct doses for HEU and for 

233U from July 1, 1972, through December 31, 1983, due to inadequate monitoring. Information is 

provided for 233U in the event that a claim is evaluated that has 233U isotopic results.

Separation of 233U from thorium occurred in three distinct campaigns: a small process test in 1965 and 

two production runs in 1966 and 1970. A total of 820 kg of 233U were produced [Walser 1978; Isochem 

1967]. Section 2.3.6 of the Hanford Site Description [ORAUT 2010a] contains a discussion of the 

program to create 233U by irradiation of thorium.

The 233U was recovered at PUREX in the nitrate form, so it would have been absorption type F. Some 

of the 233U from the 1965 run was converted to U3O8 at Z Plant [Fullam 1965, pp. 5–6; Hopkins 1965, 

p. 70], which is considered absorption type S. For Z Plant workers, because the material was being 

converted to oxide, dose reconstructors should apply absorption type F, M, or S, whichever is most 

favorable to the claimant. The Hanford bioassay database shows that a few urinalyses were obtained 

for 233U, mostly in 1970. No information has been found about how the analyses were performed. A 
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November 1970 letter from Jech to Lardy at UST indicates that aliquots from plutonium samples were 

used with a detection limit of “0.5 d/m per 100 ml” [Jech 1970b]. Copies of an earlier database show 

the units to be micrograms per liter, which is consistent with the units for all uranium analyses at that 

time. However, with the exception of a single value in 1969, the recorded values are much too small 

to be realistically in micrograms per liter. The recorded values appear to be similar to plutonium alpha 

measurements at that time, and it is probable but not verified yet that autoradiography was used for 

these bioassays because of the high specific activity of the 233U and the major contaminant 232U [50].

Based on the Jech letter, the units are most likely microcuries per sample. A second November 10 

letter from Jech indicates that the workers who were sampled were the PUREX workers involved in 

the 233U campaign [Jech 1970c]. 

There were impurities in the 233U. The measured impurities varied between the three campaigns as 

did the impurity concentrations. A summary of the highest impurities concentrations is given in 

Table 5-26. 

Table 5-26. Highest level of impurities as measured for the three campaigns. 

Impurity 1965 campaigna 1966 campaignb 1970 campaignc

U-232 ppm 2.5 6.9 8.6

Pu μg/μg U 0.002 1.8E-06d Not given

Th μg/μg U 0.094 2.2E-04 Not given

Np μg/μg U Not given 2.8E-05 Not given

Zr/Nb-95 pCi/μg U Not given 0.35 Not given

Ru-103/Ru-106 pCi/μg U Not given 0.046 Not given

Pa-233 pCi/μg U 102 4.4 Not given

Th-228 pCi/μg U Not given Not given 0.84

a. As measured at Z Plant [Fullam 1965, p. 24].

b. Isochem [1967, p. 109]. 

c. Jackson and Walser [1977, p. 177]. 

d. A very small batch had a higher concentration of plutonium.

The impurity values were converted to units of picocuries per microgram of uranium using the 

assumptions that plutonium was 239Pu, thorium was 232Th, and neptunium was 237Np with specific 

activities of 6.2 × 104 pCi/μg 239Pu, 0.109 pCi/μg 232Th, 710 pCi/μg 237Np, and 2.14 × 107 pCi/μg 232U.

Using the highest of the measured values of all the campaigns (except for the plutonium value of one 

small batch in the 1966 campaign), the results are shown in Table 5-27. These impurity 

concentrations are favorable to the claimant and are recommended for the dose reconstruction if the 

intake of 233U is determined. When determining intakes and dose for 233U intakes, determine the 

intake of 233U first, then add intakes of the impurities (including 232U); use 233U and 232U, not 234U, for 

dose calculations.

Table 5-27. Highest impurity concentrations in 233U (pCi/μg U). 

Impurity Concentration

U-232 190

Pu-239 124

Th-232 1.0E-02

Np-237 2.0E-02

Zr/Nb-95 3.5E-01

Ru-106 4.6E-02

Pa-233 102

Th-228 8.4E-01

There was exposure to HEU at Hanford, starting in July 1948, occurring frequently in 1949 through 

1956 and sporadically after that. Almost all the known work was done in the 300 Area, although there 

is indication that some work might have occurred in 231-Z. See the Hanford Site Description [ORAUT 

2010a] for details. The workers exposed to HEU were most likely also exposed to natural or slightly 
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enriched uranium. There does not seem to be a way to determine from measurements in worker files 

if an intake of uranium included HEU. If exposure to HEU seems likely, the mass urinalysis results 

should be interpreted as HEU using the 93.5% mixture in IMBA. If claim-specific evidence is found for 

HEU from July 1, 1972, through October 1983, the elemental uranium urinalysis and chest counts 

should be used to reconstruct dose. From October 1983 on, isotopic urinalysis results that indicate 

HEU (235U) should be used to reconstruct dose.

The class added by NIOSH for addition to the SEC for all employees includes HEU and 233U from July 

1, 1972, through December 31, 1983. No dose will be assigned for HEU and 233U from July 1, 1972, 

through December 31, 1983.

5.2.6 Fission Products 

Fission product urinalysis was the method for monitoring for intakes of fission products until the 

implementation of whole-body counting in 1960. Routine fission product urinalyses started in January 

1947, but ferrous hydroxide precipitation was used on the supernatant from the plutonium lanthanum 

fluoride procedure, and the results were erratic with occasional breakthrough of 40K. Therefore, data 

before 1948 should be considered unreliable and ignored.

As presented in Section 5.1.3, a class was added to the SEC for which NIOSH found that it was 

infeasible to reconstruct doses for fission products before August 31, 1946, due to inadequate 

monitoring. Therefore, dose reconstructions will not include any doses from fission products before 

August 31, 1946.

The procedure that was initiated in 1948 was to add strontium carrier to the aluminum oxide solution 

for the plutonium procedure, then precipitate lanthanum hydroxide. This procedure was shown to 

extract the rare earths and strontium with yields ranging from 90% for cerium to 23% for strontium. 

The dried planchet was counted for beta activity with an approximate detection level of 30 dpm [Healy 

1948; Wilson 1987]. The same procedure was in use in 1954 with the addition of a cerium carrier. It 

was also listed in the Old Bioassay Bible [Bioassay procedures 1961], but that same compilation has 

a separate procedure for 90Sr in urine. Another memorandum in the Old Bioassay Bible discusses the 

start of use of a gas-flow beta proportional counter in November 1958, which resulted in increased 

counting efficiency. The new detection limit was stated as 1.4 × 10-5 μCi/sample based on the 

counting efficiency of 90Sr [Bioassay procedures 1961]. “Gross fission products” are also mentioned 

by Lardy [1970] with a brief description that seems to imply the same procedure was still available, 

although probably not used much. The detection level was given as 5 dpm/sample based on the beta 

counting efficiency for 90Sr. Table 5-28 summarizes the most accurate detection levels for the fission 

product urinalysis found to date.

Table 5-28. Routine fission product urinalysis 

detection levels (dpm/sample) [51]. 

Period MDA

01/1948–02/1956 60a

03/1956–10/1964 70b

01/1965–≈12/1969 31

≈12/1970c 5

a. Reported as 30 dpm/sample, but that value was more of a 

decision level than an MDA.

b. Recorded as 3.1 or 3.17 × 10-5 μCi/sample.

c. Listed in the bioassay contract but probably not used; 

replaced by whole-body counting and 90Sr urinalyses.

It is a challenge to interpret the fission product urinalysis in a way that is meaningful as representative 

of all the possible fission and activation products to which a worker could have been exposed. The 
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procedure separated and counted radionuclides of alkaline and rare earths such as strontium, yttrium, 

barium, lanthanum, cerium, europium, promethium, zirconium, and niobium. It did not account for

radionuclides of ruthenium, cesium, zinc, cobalt, or manganese. The radiochemical yield of the 

elements that were carried through to the final planchet varied, however, from about 23% for 

strontium to about 90% for cerium [Healy 1948]. The abundances of all the fission products, relative to 

each other, varied considerably as a function of the time from when the reactor fuel was removed 

from the core and allowed to cool to when the contamination was inhaled or ingested. ORAUT-OTIB0054, Fission and Activation Product Assignment for Internal Dose-Related Gross Beta and Gross 

Gamma Analyses, provides guidance on how to interpret fission product urinalysis results [ORAUT 

2015]. Because the fission product urinalysis did not account for radiocesium, 90Sr should be used for 

the urine-based indicator radionuclide as part of the application of ORAUT-OTIB-0054. Section 5.4.1

contains a discussion about interpreting fission product mixtures.

After whole-body counting came into routine use, regular use of the fission product urinalysis 

continued for many workers at facilities such as B Plant and the Strontium Semi-Works, where intakes 

of pure 90Sr were possible. Therefore, it was apparently being used as a 90Sr bioassay. The records 

show the use of fission product analysis in this way until early 1964. The same workers show actual 

90Sr analysis results starting in 1965, probably starting with the new contract with UST. For 1960 

through 1964, fission product urinalysis results can be interpreted as 90Sr [52].

5.2.7 Strontium

Records of 90Sr urinalyses, both routines and specials, begin to appear in the database in 1965. 

However, the compilation of procedures called the Old Bioassay Bible [Bioassay procedures 1961] 

had a procedure specific for strontium in urine and fecal salts that included counting of total strontium 

and then allowing for 90Y ingrowth, yttrium separation, and counting of 90Y to account for 90Sr 

separately from gross strontium beta, if desired. This procedure was also mentioned in a July 1963 

memorandum about discussions between the Analytical Laboratories and Internal Dosimetry clarifying 

logistics of handling these samples and reporting 90Sr results. Handwritten notes on this memorandum 

indicate that the detection level was about 20 dpm. Nevertheless, the database records show the use 

of fission product urinalysis into 1964 and 90Sr urinalysis apparently starting in 1965. The database 

contains frequent entries of 1.67 × 10-5 μCi/L (37 dpm/L) during 1965 and 1966, which seems to be 

the reporting level. This is consistent with a draft of the first contract with UST (the official version has 

not been found) from August 1964 that lists a detection limit for 90Sr as 25 pCi/1.5 L, which converts to 

56 dpm/1.5L or 37 dpm/L [AEC 1964]. Lardy [1970] states that the detection limit was 1 pCi/L 

(2.2 dpm/L) (at 90% confidence) and describes the procedure as precipitation as the oxalate, then 

nitrate, removal of yttrium and barium, then reprecipitation as the carbonate and gross-beta counting 

on gas flow proportional counters. A 1974 comment about terms of the statement of work with UST 

shows an “analytical limit” (defined as ±25%) at 50 dpm/sample and a reporting level of 2 dpm/sample 

[Heid 1974]. These values appear again in the 1978 statement of work [DOE 1979] except the 

analytical limit is defined as ±100%. Robinson [1979], in a letter to R. B. Swoboda (UST bioassay 

supervisor), requests changes for 90Sr urinalyses to lower the analytical limit (±100%) from 50 to 

5 dpm/sample, increase the reporting level from 2 to 5 dpm/sample, and add an emergency analysis 

capability with an analytical limit of 10 dpm/sample and reporting level of 5 dpm/sample. In 1982, the 

detection limit was listed as 2.5 dpm/sample for 90Sr and 5 dpm/sample for 89Sr. The contract starting 

in October 1983 [Simpson 1983] lists the detection limit as 2.0 dpm/sample, and it stayed at that value 

until 1992 when it was raised to 10 dpm/sample [Acker 1992]. However, the procedure stayed the 

same throughout this period, and the true MDA probably held at about 2 dpm/sample.

Beginning in 1982, actual analytical results were recorded in the database; that is, results were not 

truncated at a reporting level.
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The results of the 90Sr procedure usually were reported as 90Sr although sometimes a value for 89Sr 

was also reported. Sometime in the 1980s a shortcut was added to the procedure that allowed 

skipping the 90Y ingrowth portion of the procedure if the first beta count was less than 1 dpm [53].

When this happened, the result was reported as Sr total or SRTOT, but the result can be interpreted 

as 90Sr. These results were below the required detection level. Table 5-29 summarizes the routine 

urinalysis detection levels for 90Sr procedure. 

Table 5-29. Routine 90Sr urinalysis detection levels [54]. 

Period MDA or MDC

Before 1965 Might have been available but MDA not known

01/1965–12/1969 1.67E-05 μCi/L (37 dpm/L)

01/1970–12/1974 1.00E-06 μCi/L (2.2 dpm/L)

01/1975–03/1979 50 dpm/samplea

04/1979–12/1981 5 dpm/sample

01/1982–09/1983 2.5 dpm/sample

10/1983–06/1990 2 dpm/sample

09/1990–11/1991 30 dpm/sample

11/1991–present 10 dpm/sampleb

a. Based on an unusual definition of analytical limit and probably 

conservative on the high side. Results <2 dpm were reported as 2 dpm; 

results >2 dpm were reported as measured.

b. Decision level was 5 dpm/sample. Before that time, the MDA was also 

used as the decision level.

Dose reconstructors should consider all strontium at Hanford to be absorption type F [55]. It is 

favorable to claimants to assume that 90Sr and total radiostrontium results are 90Sr even though 89Sr 

might be present.

5.2.8 Promethium

Hanford was involved in the manufacture of heat sources using 147Pm. The Hanford Site Description 

contains the history of possible exposure to pure 147Pm [ORAUT 2010a]. Between 500 and 600 

bioassay samples were obtained from 1966 through 1979, mostly in the 1960s. The database shows 

only seven 147Pm urinalyses from 1972 to 1975 but an upswing in the number of analyses from 1976 

to 1979, so perhaps the project ended and restarted in 1976 for a few years. In addition, animal 

studies were conducted with 147Pm as part of research to develop a human biokinetic model for the 

behavior of promethium in the body, so animal tenders might have been exposed to low levels of 

147Pm. A small study of 14 human volunteers using 143Pm was conducted in 1967 or 1968 [Palmer et 

al. 1969]. Twelve volunteers received an injection of 0.1 μCi, and two volunteers received an injection

of 10 μCi. Another study on the behavior of promethium used swine with 148mPm as the surrogate for 

147Pm [McConnon et al. ca. 1971]. 

The work on the heat sources involved conversion of promethium and cerium nitrates into Pm2O3 by 

separation chemistry then calcining [Howell and King 1968]. In addition, there was one mention of 

cold-pressed, sintered Pm2O3 for heart implants. According to ICRP [1994b], the nitrate form should 

be considered absorption type M and the oxide form absorption type S.

The units in the database are microcuries per liter for urine and microcuries per kilogram for feces 

through 1974, then disintegrations per minute per sample from 1975 to the present. Table 5-30 lists 

the 147Pm MDAs or minimum detectable concentrations (MDCs) at various times.
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Table 5-30. Routine 147Pm urinalysis detection levels [58]. 

Period MDA or MDC

01/1963–12/1964 18 dpm/L

(2.00E-06 μCi/250 mLa)

01/1965–03/25/1970 37 dpm/L

(1.67E-05 μCi/L)

03/31/1970–12/1974 22 dpm/L

(1.00E-05 μCi/L)

01/1975–12/1979 50 dpm/sampleb

01/1980–12/1981 20 dpm/sample

01/1982–09/1983 5 dpm/sample

10/1983–06/1990 4 dpm/sample

11/1991–presentc 30 dpm/sample

a. Ludwick [1964, pp. 135–138].

b. Based on an unusual definition of analytical limit and probably 

conservative on the high side. Results <25 dpm were reported as 

25 dpm; results >25 dpm were reported as measured.

c. No 147Pm analyses were performed during 1990 and 1991.

Fecal samples were analyzed for 147Pm for some of the potential intake events in the late 1960s. The 

MDA or at least the lowest reporting level appears to be 1.67 × 10-5 μCi/kg [56]. An MDA for fecal 

samples does not appear in laboratory statements of work during the 1970s, but reappears in the

1980s as 28 to 110 dpm/sample in 1982 dependent on sample size (roughly 400 dpm/kg); the value 

was 220 dpm/sample from 1983 to the 1990s [57].

The first promethium urinalysis development was in 1962 [Perkins 1963; GE 1963b, pp. 114–116]. It 

appeared to be under development in late 1962 and finished in February 1963. This procedure 

extracted rare earths and was followed by a beta count. If sufficient beta activity was detected, then 

the 147Pm component was determined by shielding to separate high-energy betas from the low-energy 

betas from 147Pm [Perkins 1963]. This procedure was used to analyze samples from the 29 workers 

who were involved in the February 12, 1963, contamination spread in the 325 Building [AEC 1963, 

Appendix D]. 

Another urinalysis procedure that used liquid scintillation counting was developed in 1963 and 

presented in the 1963 annual report of the Radiological Sciences Group [Ludwick 1964, pp. 135–138]. 

This method was described as being more amenable to routine processing. The stated detection level 

was 2 × 10-6 μCi for a 250-mL sample, which is close to the reporting level for the routine program 

later in the 1960s of 1.7 × 10-5 μCi/L.

The same description of the routine procedure that was used by UST appeared in Lardy [1970] and 

UST [1974]. Promethium and rare earths were precipitated as the fluoride. Interferences such as 

zirconium, scandium, and IV actinides were removed by extraction by TTA in xylene, first at a pH of 

less than 1, then at a pH of about 4. The final sample was counted by liquid scintillation. Remaining 

rare earths were distinguished from 147Pm by proper setting of the counting window on the liquid 

scintillation spectrometer.

In vivo counting based on bremsstrahlung from 147Pm betas was developed with an MDA of 0.5 to 

1.5 nCi [Wilson 1987, p. 84] but was not used extensively [59].

5.2.9 Polonium

As presented in Section 5.1.3, a class was added to the SEC for all employees from October 1, 1943, 

to June 30, 1972. NIOSH found that it was infeasible to reconstruct doses for polonium, neptunium, 

and thorium due to inadequate monitoring [NIOSH 2009]. The Hanford Site Description [ORAUT

2010a] contains a historical summary of work with polonium.
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A urinalysis procedure was developed in 1948 apparently in response to an accident involving a PoBe 

source in which [redacted] workers had potential for intake of 210Po [Thorburn 1948]. The procedure 

was described as wet ashing then solubilization in 2N HCl with 5 g of hydroxylamine hydrochloride 

and soaking with a silver foil for 48 hours. The polonium was attached to the silver foil, dried, and 

counted. A detection limit was not given. The control blank was 0.12 ± 0.2 dpm. The letter does not 

state the sample size but seems to imply that the results were disintegrations per minute per liter.

That would imply an MDA of about 1 dpm/L. No other records of 210Po urinalysis have been found, so 

it might have been only rarely used before 1968. The 1954 bioassay annual report stated that the 

bioassay laboratory was performing routine analyses for polonium on a few individuals [Watters 

1954, p. 93]. 

Considerable activity toward reinitiating a bioassay procedure and establishing a biokinetic model for 

210Po was found in the files from about 1968 through the mid-1970s. UST was asked to develop a 

bioassay procedure in March 1968 and did so shortly thereafter.

The procedure UST developed for 210Po in March 1968 was as follows. For urine, gold, mercury, 

platinum, and tellurium were removed by reduction in hydrazine in an HCl solution. Iron was removed 

by reduction with ascorbic acid. The polonium was then removed from solution by deposition on silver 

film by heating at 95°C for 2 hours. The silver film was counted by alpha proportional counting. Fecal 

samples were first wet-ashed in concentrated nitric acid and peroxide then treated the same as urine 

samples. Sometime between 1968 and 1974, copper foil replaced silver foil and alpha spectrometry 

counting replaced proportional counting [UST 1974]. Detection limits for routine urinalysis are shown 

in Table 5-31 and for nonroutine bioassay in Tables 5-32 and 5-33 for fecal and urine samples, 

respectively.

Table 5-31. Routinea 210Po urinalysis detection levels [60]. 

Period MDA or MDC

03/1968–12/1970 1.2 dpm/L (5.4E-7 μCi/L)

01/1971–12/1974 1.1 dpm/L (5.0E-7 μCi/L)

01/1975–12/1979 1 dpm/sampleb

01/1980–09/1983 0.1 dpm/sampleb

10/1983–present No longer listed in the contract except for expedited 

or emergency samples. Probably not used.

a. At times the routine category was called “normal.”

b. Based on an unusual definition of “analytical limit” and probably conservative 

on the high side. Reporting level listed as 0.5 dpm/sample.

Table 5-32. MDAs for nonroutine 210Po fecal analysis (dpm/sample 

unless otherwise noted) [61]. 

Period Emergencya Expedite Priority

03/1968–12/1973 Not available Not available 1.2 dpm/kg

(5.4E-7 μCi/kg)

01/1974–09/1983 Not available Not available (b)

10/1983–09/1985 340 Not available Not available

10/1985–06/1989 340 100 Not available

a. At times the emergency category was called “rush.”

b. Probably available but not listed in the contract.
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Table 5-33. MDAs for nonroutine 210Po urinalysis 

(dpm/sample) [62]. 

Period Emergencya Expedite

03/1968–12/1973 Not available Not available

01/1974–09/1983 Not available Not available

10/1983–09/1985 0.8 Not available

10/1985–06/1989 0.8 0.1

a. At times the emergency category was called “rush.”

Because 210Po is a natural radionuclide from the 238U decay chain, it exists naturally in urine and 

feces. Nothing was found in the records to indicate that a study on natural excretion levels for persons 

who lived around Hanford had been conducted.

5.2.10 Neptunium

As presented in Section 5.1.3, a class was added to the SEC for all employees from October 1, 1943, 

to June 30, 1972. NIOSH found that it was infeasible to reconstruct doses for neptunium due to 

inadequate monitoring through June 30, 1972 [NIOSH 2009]. Another class was added to the SEC 

because NIOSH found that it was infeasible to reconstruct doses for neptunium, from July 1, 1972, 

through December 31, 1983. No dose will be assigned for neptunium through December 31, 1983.

At 222-S (1958 through 1962) and the PUREX Plant (1958 through June 1972), 237Np was removed 

from the dissolved fuel, purified, and packaged for shipment off the site. Until the final purification 

step, the material had more activity of plutonium than neptunium, but the final product was very pure.

Neptunium target elements, which used the purified neptunium, were manufactured in 231-Z and in 

the 308 and 3708 Buildings intermittently from 1958 to 1967. See the Hanford Site Description for 

details [ORAUT 2010a]. 

A urinalysis method for 237Np was developed in 1960. There is no evidence in the database that 

suggests this was used until 1972, but hard-copy results might show up in workers’ files for earlier 

times. The procedure used neptunium separation followed by electrodeposition and alpha track 

counting in the same manner as for plutonium urinalysis samples, with essentially the same sensitivity 

(0.05 dpm/24-hr sample) [Perkins 1961, p. 6]. 

The database contains only four bioassay samples for 237Np in the 1970s for PNNL workers; all four 

are baseline samples. No routine samples or ending work samples were collected. The value for all 

four samples was 1 × 10-7 with no units shown. The practice at the time was to record the activity of 

excreta samples in units of microcuries, but it has not been determined to date if the units are per 

sample or per liter. No information about the analysis method has been found [63].

A few sample results appear in the database in the 1980s, mostly for PNNL employees. The units are 

in disintegrations per minute per sample [64]. The 1983 statement of work with UST [Simpson 1983] 

shows a nonroutine analysis for 237Np with contractual detection levels of 0.02 dpm/sample for urine 

samples and 0.1 dpm/sample for fecal samples, which were consistent with alpha spectrometry for 

plutonium and americium at the time.

5.2.11 Other Limited-Exposure Radionuclides

Hanford has always been a center for research, first as the Hanford Engineer Works for the 

Manhattan Project through 1946, then as Hanford Works, then (1965 to present) at PNNL. As such, 

small-scale use (in terms of either the number of persons or activity of the source) of various 

radionuclides that are not addressed above has occurred throughout the history of the Hanford Site.

The following discussion addresses 14C, 232Th, radon, 90Y, 227Th, 227Ac, and 32P, but it is probably not 

comprehensive.
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Carbon-14

Exposure occurred at the 3731 Building in the mid-1950s when irradiated graphite samples were 

brought to the building from the operating reactors for destructive testing. No information has been 

uncovered about what bioassay, if any, was done. Carbon-14 was also used as a tracer in biological 

experiments. One documented study was conducted in the late 1990s in the Life Sciences 

Laboratory-II (LSL-II) that involved a total of about 4 Ci of 14C. Urinalyses were obtained on about 

20 researchers. The MDA was 10 dpm/mL [DOE 1987; Lyon et al. 1991; Carbaugh 1997]. Baseline 

samples were obtained from each worker because natural excretion levels had not been established.

ICRP [1994b, 1995] assigns 14C in organic compounds to class SR2.

Thorium

As presented in Section 5.1.3, a class was added to the SEC for all employees from October 1, 1943, 

to June 30, 1972. NIOSH found that it was infeasible to reconstruct doses for thorium due to 

inadequate monitoring through June 30, 1972 [NIOSH 2009]. Another class was added to the SEC 

because NIOSH found that it was infeasible to reconstruct doses for thorium, from July 1, 1972, 

through December 31, 1983. No dose will be assigned for thorium through December 31, 1983.

Thorium urinalysis capability apparently existed at least as far back as 1955 [Nickelson et al. 1955, 

pp. 24–25]. However, there is no indication in the database of thorium samples until 1979. The 1955 

monthly report indicates a detection limit of 0.077 μg but does not state sample size. There are some 

thorium urinalysis samples in the database beginning in 1979. These are identified as either TH or 

Th232. The TH urinalysis was a total thorium analysis by spectrophotometry, with a detection level of 

1 μg/L. The TH232 analysis was specific for 232Th and used alpha spectrometry. The detection levels 

were listed as 0.5 dpm/sample in 1970 [Lardy 1970] and 0.03 dpm/sample in 1983 [Simpson 1983]. 

Some work was done with 232Th slurries in the 3720 Building in the mid-1990s. The plan was to collect 

baseline urine samples on the few involved workers, then collect special bioassay samples if air 

samples exceeded a cumulative exposure of 40 derived air concentration (DAC)-hr. The urinalysis 

MDA was stated to be 0.1 dpm/sample [Bihl 1994; Lyon et al. 1998]. 

Sulphur-35

There are some urinalysis results for 35S in 1967. According to Lardy [1970], analysis was

accomplished by ashing then liquid scintillation counting with a detection limit of 10 dpm/sample.

However, as with 14C, the results were reported as microcuries per liter and the reporting level was 

apparently 1 μCi/L.

Radon

A radon generator was used for animal studies in the 108F Building; it was later moved to LSL-II.

Monitoring was probably just by air sampling, but no information has been discovered yet. The radon 

generator was housed in a glovebox and vented to the outside. There should have been only a few 

potentially exposed researchers [65].

Phosphorous-32

This radionuclide was used for biological tracer studies and, according to one retired researcher, 

“pipetting was done by mouth in the old days” [66]. Such exposure would be limited to a few 

researchers and would have to be established through the claimant interview or by some indication of 

32P bioassay samples in the worker’s record. More information might be found if such a case is 

encountered. The whole-body counter was calibrated for 32P measurements in about 1965, and there 

might be personnel count results in the records [Keene 1965, p. 6]. 

In addition, some unusual radionuclides were isolated in the 325 Building for nuclear medicine studies 

in the mid to late 1990s: 
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• Yttrium-90. One of these studies isolated 90Y from 90Sr and packaged and shipped the 90Y to 

users around the world. Only a few workers were involved. The work was monitored by air 

samplers and no loss of control of the material occurred, so no bioassay was obtained. The 

material was in an insoluble form, so chest counting would have been the only possible 

bioassay because of the 64-hour physical half-life, but the need to perform chest counting 

never arose [Carbaugh 1990; Bauman 1996]. 

• Thorium- and Actinium-227. Another project in about 1995 to 1996 involved milking 227Th from 

227Ac on an ion exchange column. A bioassay procedure was developed specifically for this 

project under the assumption that the project was going to continue for several years, but it 

ended after only a few milkings. Only a couple of researchers were involved [Bihl 1995]. No 

contamination spreads or high air samples occurred. The bioassay procedure had a stated 

MDA of 0.1 dpm/sample for 227Th.

5.3 IN VIVO MINIMUM DETECTABLE ACTIVITIES, ANALYTICAL METHODS, AND 

REPORTING PROTOCOLS

In vivo counting equipment and techniques were developed in the late 1950s and have been in 

routine use for measuring X- and gamma-ray-emitting radionuclides since 1960. (Unless otherwise 

noted, the in vivo information below is from Wilson [1987] and Lynch et al. [2001].) The unit for all in 

vivo measurements is nanocuries.

5.3.1 Whole-Body Counters

The first whole-body counting started in mid-1959, and it became a routine method in 1960. It 

consisted of a single NaI crystal (9.375 in. diameter and 4 in. thick) in a counting room with 10-in.-

thick pre-World War II steel plates on all six sides and graded shielding on the inner surfaces (lead, 

cadmium, and copper) [Wilson 1987; Roesch et al. 1960]. This was called the Iron Room. The 

counting geometry was a chair that was configured to simulate a 1-m arc. The original count time was 

20 minutes, which was reduced to 10 minutes in October 1962 [Swanberg 1962]. A second, samesize NaI detector was added in 1963 [Brady 1964]. According to the personal recollection of H. E. 

Palmer, the two-detector system improved the detection capabilities somewhat. However, the MDAs 

in a report from the fall of 1964 were the same as those in Palmer’s Laboratory Record Notebook in 

1960 [67], so the difference between the systems was apparently not great enough to warrant 

republishing the MDAs. Therefore, the MDAs in Table 5-34 are the only MDAs found for the 1960s 

and 1970s, and they apparently were meant to apply generally to the various whole-body counters in 

operation during this period. Although they were referred to as detection limits, the calculations were 

more consistent with what is presently referred to as decision levels.

Table 5-34. Routine whole-body counting detection levels (nCi).

a

01/1960–12/1976b

Nuclide MDA Decision level Reporting level

Na-22 Not applicable 1.0 10

Na-24 Not applicable 0.3 0.3

Cr-51 Not applicable 50 50

Fe-59 Not applicable 2.0 10

Co-60 Not applicable 0.4 10

I-131 Not applicable 0.5c 10

Cs-137 Not applicable 0.5 0.5

Zn-65 Not applicable 1.6 0.8

01/1977–12/1984d
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Nuclide MDA Decision level Reporting level

Na-22 Not applicable 1.0 10

Na-24 Not applicable 0.5 0.31

Cr-51 Not applicable 15 15

Mn-54 Not applicable 2.0 10

Fe-59 Not applicable 4.0 10

Co-60 Not applicable 2.0 10

Zn-65 Not applicable 3.0 0.75

Zr/Nb-95 Not applicable 2.0 10

Ag-110m Not applicable 2.0 10

Ru-106 Not applicable 12 12

Sb-125 Not applicable 3.0 10

I-131 Not applicable 4.0c 10

Cs-137 Not applicable 2 0.66e

Ce-144 Not applicable 100 100

01/1985–12/1986

Nuclide MDA Decision level Reporting level

Na-22 Not applicable 1 1.5

Na-24 Not applicable 1 0.31

Mn-54 Not applicable 1 3

Fe-59 Not applicable 2 6

Co-60 Not applicable 1 1

Zn-65 Not applicable 2 0.75

Zr-95 3 1 3

Ru-106 12 4 12

Eu-154 4.5 5 4.5

Cs-137 3 1 0.66

01/1987–12/1987f

Nuclide MDA Decision level Reporting level

Na-22 1.5 Not applicable Anything detected

Mn-54 3 Not applicable 6

Co-60 3 Not applicable 5

Fe-59 6 Not applicable Anything detected

Cs-137 3 Not applicable 6

01/1992–12/1992g

New formalism for decision level calculation; “limit” in database changed from 

MDA to decision level.

01/1993–12/1994h

Actual values, regardless of amount, reported for Co-60 and Cs-137, including 

negative numbers.

01/1995–10/1999i

Nuclide MDA Decision level Reporting level

Co-60 4 Not applicable Every result

Cs-137 4 Not applicable Every result

I-131 5 Not applicable Every result

Mn-54 3 Not applicable Every result

Na-22 2 Not applicable Every result

Na-24 1 Not applicable Every result

Pr-144 (Ce-144) 230 Not applicable Every result

Other radionuclides Not applicable Not applicable Anything detected
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10/1999–presenti,j

Nuclide MDA Decision level Reporting level

Co-60 1.25 (g) Every result

Cs-137 1.3 (g) Every result

Eu-154 3.75 (g) Every result

Other radionuclides Not applicable Not applicable Anything detected

a. Nominal MDAs based on the phantom available at the time, the routine count time, and 

the least sensitive of various whole-body counters in operation at the time. Listing of an 

MDA for a given radionuclide does not necessarily mean that that radionuclide was 

frequently encountered. If smaller MDAs are listed in the database for a given count, 

use them.

b. Based on 95% confidence of detection [68]. There were some special counts for 

thorium; see Section 5.3.4. 

c. See discussion on thyroid detectors.

d. Based on 99% confidence of detection [Wilson 1987]. 

e. The reporting level cutoff of 0.66 nCi for Cs-137 appears to start in 1975 [69].

f. From Lyon et al. [1988]. No changes for other radionuclides. Anything detected is 

reported.

g. From Lyon et al. [1993]. 

h. [70].

i. Least sensitive of many options during the period. Much better sensitivities were 

available using the hyperpure germanium (HPGe) system [BMI 1995, Tables 7.2 and 

7.3]. 

j. Physical configurations stayed essentially the same but ABACOS software introduced 

changes to the methodology for determining MDAs and decision levels. Decision levels 

were determined by software on a count-by-count basis [Lynch et al. 2000]. 

Shortly after the chair counter in the Iron Room became operational, an entirely new design called the 

shadow shield counter was developed. The shadow shield consisted of a bed shielded on the bottom 

and sides by lead. The bed moved under a large NaI crystal (11.5 in. diameter by 4 in. thick) that was 

also shielded by lead with the exception of the downward-looking face that looked directly onto the 

body as it passed under the crystal. The shadow shield detector was mounted in a mobile trailer and 

moved to areas nearer the Hanford worksites. The trailer also had a thyroid detector and a wound 

counter. The shadow shield detector became operational in 1963 [Brady 1964]. The mobile counter 

was described as having comparable sensitivity to the “larger, conventional whole-body counters 

installed in massive iron rooms. There is, however, some decreased sensitivity in the lower energy 

region below about 300 keV, due to increased contribution to the background from scattered 

radiation” [Swanberg 1963]. 

A report on the radionuclides detected in workers at the whole-body counter facility in 1961 listed 

24Na, 60Co, 65Zn, 95Zr, 95Nb, 99Mo, 99Tc (presumably 99Mo), 103Ru, 106Ru, 131I, 137Cs, and 144Ce [Henle 

1962]. A similar report that summarized 1961 to 1963 results added 46Sc, 51Cr, and 59Fe to the list 

[Brady 1964]. 

A shadow shield whole-body detector was added at the Whole Body Counting Facility in 1977. This 

assembly had two 35% lithium-drifted germanium (GeLi) detectors and a 4- by 4- by 16-in. NaI 

detector. It ceased operation in 1987 when the two new counting rooms were added. Wilson [1987] 

provided a list of MDAs that applied to 1980. These are used to represent this shadow shield detector.

By 1978, there were four shadow shield whole-body counters available for use: one at the Whole 

Body Counting Facility, two in mobile trailers, and one at the Emergency Decontamination Facility.

The latter was designated for use for large acute intakes with potentially high levels of external 

contamination.

A “standup” counter put in operation in 1983 is still in operation today. It consists of five vertically 

stacked NaI crystals in a small lead-shielded area. The worker stands in front of the detectors with the 

detectors to his or her back; the detector array is raised or lowered to best fit the height of the worker.
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There are four 9.375-in.-diameter by 4-in.-thick detectors and one 11-in.-diameter by 4-in.-thick 

detector; the latter is located behind the thoracic region. The count time is 200 seconds [Palmer et al.

ca. 1984].

A coaxial HPGe scanning array developed in July 1989 also still operates today [Lyon et al. 1990]. For 

this system the person lies on a bed in a shielded room and the detector array moves under the bed.

The configuration of this system, in terms of number and size of the detectors, has changed many 

times. It started as four 68% HPGe detectors; one of the detectors was replaced with a 120% detector 

in late 1995; in May 1997 the system was upgraded to include seven detectors including three 120% 

detectors. As a four-detector array, the system was used only when the standup counter revealed

detectable activity of an occupationally related radionuclide. However, it was considered the count of 

record. In 1997, because of its greater resolution and lower decision levels, it began to be used for 

routine counts for workers who were exposed to mixtures of 137Cs and plutonium. The count time is 

usually 10 minutes, but 20-minute count times are used as confirmation of an initial count with 

detectable activity. As a consequence, the database usually shows both 10- and 20-minute counts on 

the same day or a few days apart if the first count had detectable activity (excluding 40K or medical 

radionuclides) [Lyon et al. 1998].

The radon progeny 214Bi and the thoron progeny 208Tl were and are present in the background spectra 

from the shielded rooms used for in vivo measurements of Hanford workers. To account for this 

contribution when calculating worker results, the background levels of these nuclides have been 

estimated from background measurements in the rooms with detectors shielded with an appropriate 

phantom. The environmental background count rates, in addition to the continuum background count 

rates, are subtracted from the gross count rates in the appropriate region of interest of the worker 

spectra to obtain an activity for these nuclides. This approach was implemented in May 1992 [Lynch 

1992].

Even with the background contribution subtracted, some periodically measured activity levels of 214Bi 

were higher than room background levels due to activity on the workers’ clothes, skin, and hair. To 

compensate for this, derived investigation levels (DILs) were implemented in October 1995 for 214Bi 

and 208Tl measurements from the coaxial germanium counting system. The DIL values were 6.47 nCi 

for 214Bi and 0.75 nCi for 208Tl, which represent the upper 99% confidence levels for an unexposed 

population. Results were considered to be above detection and were reported to the REX database 

only if the value exceeded the DIL value. The use of the DIL values continued until November 1999 

when a new software system was implemented. Since 1999, 214Bi and 208Tl results have not been 

reported unless specifically requested.

Use of the first mobile counter at onsite locations stopped in the early 1980s. A new trailer was 

obtained in 1989 and reconfigured with a new standup counter consisting of five 4- by 16- by 4-in. 

thick NaI detectors plus one 4- by 8- by 4-in. thick detector. The trailer was parked in the 200 East 

Area and operated remotely starting in 1991 [Lyon et al. 1992]. The sensitivity of the detector was 

comparable to that of the standup counter at the Whole Body Counter Facility. The facility was used 

infrequently and its use was discontinued in August 1995 [Lyon et al. 1996].

From 1960 to 1983, four radionuclides were reported routinely: 24Na, 40K, 65Zn, and 137Cs [71].

Potassium-40 is strictly a natural source and can be ignored. Sodium-24 and 65Zn were detectable in 

many whole-body counts in the 1960s. Most of these intakes came from sanitary water drawn from 

the Columbia River. The onsite reactors and many of the cities around the Hanford Site used the 

Columbia River for sanitary water, although the 24Na intakes came from the drinking water at the 

reactors and 200 Area (which received water from the 100 Area) because its short physical half-life 

greatly reduced the activity in city water supplies [Brady 1964]. Therefore, intakes of 24Na and 65Zn 

can be considered chronic ingestions, with the exception of the highest body burdens in Hanford 

workers, which were statistically different from the general background and had to come from 
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inhalations. Guidance to distinguish acute inhalations from chronic ingestions based strictly on the 

measured activity level has not been developed. Therefore, barring evidence in the employee’s file 

that ties the intake to an acute event, the dose reconstructor can assume either chronic ingestion or 

inhalation in accordance with which is more favorable to the claimant for the organ of concern.

Chronic ingestion intakes would have occurred before the implementation of whole-body counting, 

probably from shortly after the startup of the first once-through-cooling reactor (August 1944), and 

they would have ceased shortly after the shutdown of the 100 KE reactor (January 1971) [72].

Net counts in a fifth region of the spectrum were commonly calculated but not usually associated with 

a radionuclide. This was the low-energy portion of the spectrum noted as the GOK region. The 

technique was to calculate the activity of the higher-energy radionuclides 24Na, 40K, etc., then subtract 

the Compton scatter contribution from those radionuclides to see if any counts were left over in the 

low-energy region. If sufficient counts were left over, they would have been investigated further to see 

if an occupational radionuclide was the source, recognizing that the low-energy region was also 

subject to increased electronic noise and general background scatter in the crystal [73].

If the hard-copy form (In-Vivo Counter Results) shows the “traces of xxx invalidate routine calculation” 

statement, some radionuclide other than the standard four was detected; this was often 60Co. The 

activity of that radionuclide might be written on the form. Activities that exceeded 10 nCi or 1% of the 

MPBB were calculated and reported on a Whole Body Counter Evaluation form [Glenn 1968]. See 

Section 5.3.6 for instructions.

Most workers in the early days of whole-body counting had detectable activities of 137Cs. Most of this 

was attributed to fallout. Some workers had even higher levels of 137Cs from consumption of wild 

game. A decision level to establish the difference between occupational and nonoccupational sources 

of 137Cs intake has not been found in the records, and might not have been developed as long as the 

137Cs measurement did not exceed 1% of an MPBB. Dose reconstructors can use the following 

guidance [74]:

• Consider the 137Cs intake occupational if the same whole-body count detected other fission or 

activation products (excluding 65Zn or 24Na). Also consider it occupational if a fission product or 

radiostrontium urinalysis showed detectable activity and the sample was obtained within the 

period between the previous and next whole-body counts.

• If an investigation record clearly shows that the intake was due to a nonoccupational source, 

disregard the 137Cs.

• NCRP Report 94 provides mean body burdens of 137Cs for the United States for the years 

most likely to produce interference with occupational whole-body count results [NCRP 1988]. 

Table 5-35 lists those values. If no other fission or activation products are linked to the intake 

(excluding 65Zn or 24Na) and the 137Cs result is less than the values in Table 5-35, the dose 

reconstructor may assume the 137Cs result is due to fallout.
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Table 5-35. Mean body burdens of 137Cs 

from fallout in the United States (nCi).a

Year Body burden

1953 0.27

1954 1.1

1955 2.2

1956 4.3

1957 5.1

1958 6.5

1959 8.1

1960 6.8

1961 4.6

1962 6.0

1963 11

1964 19

1965 16

1966 9.7

1967 5.6

1968 3.5

1969 2.7

1970 2.7

1971 2.7

1972 2.7

1973 2.7

1974 1.6

1975 1.1

1976 1.6

1977 1.1

a. From NCRP [1988]. 

5.3.2 Chest Counters

In 1967, PNNL started using the original large NaI detector in the Iron Room for chest counting, with 

emphasis on uranium workers. The detector was placed directly over and nearly in contact with the 

chest region with the worker in the supine position. The count time was 30 minutes. MDAs were 

determined to be 6.7 nCi for “U natural” (presumably based on 234Th), 0.15 nCi for 235U, and 0.33 nCi 

for 241Am. However, in the next year a new counting room, called the Lead Room, was built 

specifically for chest counting. It had four 5-in.-diameter by 0.375-in.-thick NaI detectors, two in front 

and two in back of the subject. The count time was 30 minutes [Wilson 1987]. A lung phantom with 

variable chest wall thickness was developed for calibration of the new system. MDAs were listed as 

0.15 to 0.6 nCi for 241Am, 2.0 to 3.7 nCi for 234Th (assumed to be in equilibrium with 238U), and 0.17 to 

0.37 nCi for 235U, dependent on a subject’s weight-to-height ratio [Andersen 1971; Wilson 1987]. 

(Chest count MDAs are summarized in Table 5-36.) MDAs for direct measurement of 238Pu and 239Pu 

using the 17-keV X-rays were calculated at times, but the values were extremely large in relation to 

the maximum permissible lung burden, so primary reliance was placed on measuring 241Am and 

applying a plutonium-to-americium ratio. The chest counter was also calibrated to measure 

bremsstrahlung radiation from 90Sr or 147Pm, although these were probably not routine counts. MDAs 

for those counts were listed as 25 to 40 nCi and 0.5 to 1.5 μCi for 90Sr and 147Pm, respectively 

[Andersen 1971]. A second chest counting system became operational in 1978 [Wilson 1987]. A 

phoswich detector became available and was used occasionally for special chest counts but was 

never implemented on a routine basis [75].
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Table 5-36. Routine chest counting detection levels (nCi) [Andersen 1971; Lyon et al. 1988, 1989,

1990, 1993, 1994; MacLellan et al. 1999; Lynch et al. 2000]. 

01/1967–12/1967

Radionuclide MDA

Am-241 0.33

U-238 (Th-234) 6.7

U-235 0.15

01/1968–1983

Radionuclide MDA

Am-241 0.15–0.6a

U-238 2.0–3.7a

U-235 0.17–0.37a

Sr-90 25–40a

Pm-147 0.5–1.5a

01/1983–12/1986

Radionuclide MDA

Am-241 0.24b

U-238 1.1b

U-235 0.08b

Ce-144 0.78b

Eu-154 0.07

01/1987–12/1987

Radionuclide MDA

Am-241 0.28c

U-238 1.8

U-235 0.12

Ce-144 0.6

Eu-154 0.07

01/1988–06/1989

Radionuclide MDA

Am-241 0.28c

U-238 1.8

U-235 0.12

Ce-144 0.6

Eu-154 0.07

07/1989–12/1991d

Radionuclide MDA

Am-241 0.18c

U-238 1.2

U-235 0.08

01/1992–05/1996e

Radionuclide MDA

Am-241 0.18c

U-238 3

U-235 0.2

06/1996–10/1999

Radionuclide MDA

Am-241 0.28c

U-238 1.6

U-235 0.095

11/1999–present 

Radionuclide MDA

Am-241 0.25c

U-238 1.5

U-235 0.090

a. Range for different weight-to-height ratios, a chest wall thickness adjustment for both front and back 

chest walls. Use highest value for default to cover large persons.

b. Assumed MDA = (RDA)(4.65/3). Am-241 adjusted for 95th-percentile male chest wall (0.2/0.13) 

[Lynch 2003].

c. Adjusted for 95th-percentile male chest wall.

d. Ce-144 and Eu-154 no longer automatically reported for chest counts because they now can be 

quantified in the germanium whole-body counter.

e. Applies to the six-detector array. Better sensitivity was obtained by the four-large-area-detector array 

in the Stainless Steel Room.

A solid-state germanium counting system using three planar high-purity germanium (HPGe) detectors 

(soon upgraded to six) replaced the NaI detector in the Iron Room chair counter in 1983 [Palmer et al.

ca. 1984]. The HPGe detectors provided better spectral resolution than the NaI detector and

consequently lower backgrounds in the region of interest, better discrimination against radon decay 

products, and better detection of low-energy photon emitters in the presence of large activities of highenergy photon emitters (e.g., 137Cs or 60Co). A thin window on the end of the detector faced the chest 

for better transmission of low-energy photons. The detectors were positioned over the front of the 

chest (two over the right lung) with the subject in the supine position. The count time was 

2,000 seconds. MDAs were quoted for “an average size person” as 0.1 nCi for 241Am, 0.5 nCi for 

144Ce, 0.7 nCi for 234Th (238U), and 0.05 nCi for 235U [Palmer and Rieksts 1984]. These values were 

quoted as being the reliably detectable activity (RDA), which was defined as 3 standard deviations of 

the background continuum and was discernable by naked-eye inspection of the spectrum [Carbaugh 

et al. 1988]. Special chest counts as followup to high routine chest counts or on special request were 

twice the normal counting time, so the MDAs were somewhat lower. 
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Within less than a year, the three-detector system was upgraded to a six-detector array, which 

enabled reduction of routine counting times to 1,000 seconds with nearly the same RDAs [Palmer et 

al. ca. 1984; Carbaugh et al. 1988]. A second HPGe detector array became operational in July 1989 

in a new shielded cell called the Stainless Steel Room because the inner (i.e., visible) lining of the 

graded shield was stainless steel. Although intended to be a six-detector array, this counter had only 

four detectors at first because of operational problems with the detectors. Count times were increased 

to 2,000 seconds for the six-detector array and 3,000 seconds for the four-detector array [Lyon et 

al. 1990]. 

In September 1994, the chest counter in the Stainless Steel Room was converted to a four-detector 

array using larger area detectors [Lyon et al. 1995]. The same change was implemented in the Iron 

Room in June 1996 [Lyon et al. 1997]. This configuration continues to the present. The routine count 

time was increased to 3,000 seconds for the larger area arrays in November 1995; special counts and 

recounts were 3,600 seconds.

Ultrasonic measurement of chest wall thickness for workers who had activity in the lung began in 

about 1978 and continues today [Palmer and Rieksts 1979]. Therefore, decision levels for 

nondetected activities use a weight-to-height ratio to estimate chest wall thickness, whereas detected 

activity is corrected for chest wall thickness using ultrasound [76].

Individual-specific decision levels were reported to the database for each count and each radionuclide 

starting in 1992 [Lyon et al. 1993]. 

For in vivo counting, the assumption was made that 234Th was in equilibrium with 238U [77]. This was a 

reasonable assumption at Hanford. Uranium that was recently separated from dissolved fuel was 

certainly not in equilibrium, and uranium being treated at the UO3 Plant might have been in equilibrium 

depending on how long it had taken the material to undergo the separation process and transport to 

the UO3 Plant. However, uranium in this part of the fuel cycle was very soluble and not important in 

relation to chest counting. Chest counts were used to monitor for intakes of insoluble forms of 

uranium, which were very old forms in terms of time since purification from decay progeny (e.g., 

machining on metal and uranium metallurgy studies) [78].

5.3.3 Thyroid Counters

Thyroid counting appears to have started on a limited basis for high-risk workers at least as early as 

1956 (see Section 5.6.1 for a discussion on thyroid counting in 1945 and 1946). Wilson [1960] states,

At the present time routine thyroid monitoring is conducted on a limited basis in the 

REDOX and PUREX facilities. Generally the pattern for coverage in the PUREX facility 

includes about four to five employees weekly picked from the sampling crews, crane 

operators, and a Radiation Monitor assigned to the stack area. At the REDOX facility 

routine monitoring is accomplished on a weekly basis for the shift crane operators.

The letter continues to discuss counts and other data that were obtained in 1959, but there is no 

indication if those results were placed in workers’ files. Radiation monitoring data sheets from 1956 

show that results below 10 nCi for 131I were recorded as less-than values.

The first mobile whole-body counter had a thyroid counter that consisted of a 3- by 3-in. NaI detector 

(assumed to mean 3-in. diameter by 3-in. thick) that was positioned next to the neck. The MDA was 

0.020 nCi for 131I for a 30-minute count [Wilson 1987]. The same detector and MDA were included in a 

description of in vivo counting capabilities at the Whole Body Counting Facility by Andersen [1971] 

and again by Palmer [ca. 1985]. In 1986, the first listing of MDAs in the database associated with 

each count occurred; because the MDA is different for each thyroid count, person-specific calculations 
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were used. There is a cluster of thyroid counts with MDAs that range from 0.020 to 0.050 nCi and 

another cluster that ranges from 0.22 to 0.298 nCi. The second cluster occurred only in October 1986 

and would seem to imply use of a different detector for some reason. The count-specific MDAs after 

October 1986 vary from 0.07 to 0.006 nCi [79].

The presence of 131I would have been detected in a whole-body count, but the recorded quantity 

would have been based on a whole-body calibration rather than a thyroid calibration [80].

It is important to note that, during some years, 131I that was measured in workers who had radioiodine 

medical studies or treatment was recorded in the database; the workers’ files should have a hardcopy in vivo count form that notes the 131I was due to a medical procedure. Dose reconstructors must 

look for such a note in the records of anyone who has a recorded 131I measurement if the 131I result 

has impact on the probability of causation. 

For counting 125I in the thyroid, a thin, 2-in.-diameter NaI crystal with a beryllium window was used 

starting at least as far back as 1967. The thickness of the crystal has not been determined yet. The 

MDA was listed as 0.11 nCi for a 1-minute count or 0.07 nCi for a 10-minute count, but there was no 

mention of which count time was regularly used. There probably were not many workers who were 

exposed to 125I on a regular basis, but there are indications of a contamination spread in 1978 that 

involved several workers [Palmer and Rieksts 1979]. The same counter is described for thyroid 

counting in 1982, except the “reporting level” is given as 0.020 nCi; it is not known if this better 

sensitivity came from a longer count time, better positioning, or an improved crystal.

By 1985, thyroid counting for 125I was performed using two intrinsic germanium detectors, with an 

MDA of 0.005 nCi for a 2,000-second count.

Thyroid counting for either of the iodine isotopes has been rare since 1987.

5.3.4 Thorium Exposure and Monitoring

As presented in Section 5.1.3, a class was added to the SEC for all employees from October 1, 1943, 

to June 30, 1972. NIOSH found that it was infeasible to reconstruct doses for thorium due to 

inadequate monitoring through June 30, 1972 [NIOSH 2009]. Another class was added to the SEC 

because NIOSH found that it was infeasible to reconstruct doses for thorium from July 1, 1972, 

through December 31, 1983. No dose will be assigned to unmonitored workers for thorium through 

December 31, 1983. Information is provided for thorium and associated radionuclides in the event that 

a claim is evaluated that has thorium urinalysis and/or in vivo bioassay results for 232Th (likely based 

on 228Ac results above the MDA).

Thorium exposure occurred at many locations throughout Hanford’s history; the Hanford Site 

Description contains details [ORAUT 2010a]. Table 5-37 lists buildings where there was a potential for 

exposure to thorium. Although a thorium urinalysis procedure was in place at least as far back as 

1955 [Nickelson et al. 1955, pp. 24–25], there are no thorium urinalysis results in the database until 

1979. Whole-body counting specific for 232Th first shows up in 1962 (one worker) and 1963 (two 

workers) with tens of counts per year in the later 1960s and a few counts in the 1970s.

Radiation protection practices were similar for thorium and uranium, so dose reconstructors should 

assume a chronic intake. Because metal, oxide, nitrate, and perhaps other forms of thorium were 

being handled, either absorption type M or S can be assumed. Radium-228, the first progeny of 232Th, 

would have been removed either during the processing of the ore at Fernald or during baking and 

sintering in the 300 Area, so the progeny would not have been in equilibrium [West 1965]. Workers 

could have been exposed to thorium in a mixture of ages (i.e., time since purification), but most of the 

material would have been fairly young. It is favorable to claimants to assume younger material when 
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whole-body counting is the bioassay method (as explained below). Therefore, the thorium was 

assumed to have aged 0.5 year since purification. Therefore, for an intake of 232Th by activity, dose 

reconstructors should assign 228Ra at 0.058 times the 232Th intake and assign 228Th at 0.84 times the 

232Th intake [81]. The 230Th specification for the thorium oxide was <1 ppm [AEC 1968, Spec. No. T100], which means that 230Th contamination would have contributed <6% of the total alpha activity and 

can be ignored.

Table 5-37. Buildings with thorium processing

[Gerber 1992; Isochem 1967; Walser 1978]. 

Building, area Date

313, 300 1945–1970

314, 300 1945–1970

305,300 1945–1946

306, 300 1955–1970

3706, 300 1954–1963

3707 A&B, 300 1945–1970

3732, 300 1965–1970

3722, 300 1946–1970

325, 300 1963a

321, 300 1964–1970

202 A, PUREX, 200 East 1965–1970

241-WR vault, 200 West 1965–1978

224-U, 200 West 1966–1967

a. Hot cell work only. Worker exposure unlikely.

Routine monitoring for thorium intakes was by whole-body counting based on 228Ac. Urinalyses were 

used for special investigations. Jech [1967–1969, p. 3] stated the detection limit for 232Th was 1.2 nCi. 

Jech [1967–1969, p. 4] summarized the results of a worker’s whole-body count and stated, “...and no 

detectable thorium (<1 nCi)”. The 1.2-nCi detection limit was probably a decision level, so the MDA 

would have been about 2.4 nCi [82]. Sometimes the whole-body count record shows 232Th in the list of 

radionuclides if the worker was specially earmarked for possible exposure to 232Th, but the dose 

reconstructor should not count on this for every exposed worker. If nothing was detected, no activity is 

shown. For a few of those workers no record of the whole-body count was found in the database, but 

a paper record of the count was submitted as part of the DOE files. Although the in vivo records show 

the results as 232Th, the measured activity was 228Ac [83]. The two radionuclides would not have been 

in equilibrium; both disequilibrium at intake and separate biological processes on the 232Th parent and 

the 228Ra progeny mean that the 232Th activity in the body would not have been the same as the 

measured 228Ac. Assuming chronic intake and exposure to 0.5-year aged mixture, the factors in 

Tables 5-38 and 5-39 (for 232Th types M and S, respectively) should be used to determine the intake 

of 232Th per nanocurie of 228Ac (reported as 232Th) that was measured in a whole-body count. Once 

the adjusted intake of 232Th has been determined, the ratios in the paragraph above must be used to 

determine the intakes of 228Ra and 228Th.
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Table 5-38. 232Thorium intakes from whole-body count measurements, 

type M [84].

Period of chronic intake 

before whole-body 

count (d)

Ratio of Th-232 to 

Ac-228 in whole body at 

time of measurement

Intake of Th-232 per 1 nCi 

of Ac-228 (reported as 

Th-232) measured in 

whole-body count (pCi/d)

182 23.17 1,810

364 26.56 1,150

546 29.03 874

728 30.40 704

910 30.73 580

1,092 30.28 483

1,274 29.34 405

1,463 28.09 341

1,645 26.79 291

1,827 25.49 251

2,009 24.24 218

2,184 23.10 192

Table 5-39. 232Thorium intakes from whole-body count measurements, 

type S [85].

Period of chronic intake 

before whole-body 

count (d)

Ratio of Th-232 to 

Ac-228 in whole body at 

time of measurement

Intake of Th-232 per 1 nCi 

of Ac-228 (reported as 

Th-232) measured in 

whole-body count (pCi/d)

182 11.33 1,310

364 9.218 648

546 7.834 412

728 6.898 297

910 6.234 232

1,092 5.743 191

1,274 5.370 162

1,463 5.068 142

1,645 4.837 127

1,827 4.649 115

2,009 4.494 106

2,184 4.369 98.6

NOTE: Use of whole-body counting for thorium monitoring was apparently different at other 

AEC sites that used chest counting. Be sure to recognize that the following discussion and 

Tables 5-38 to 5-41 apply only to whole-body counts.

The dose reconstructor can use the whole-body count record with the urinalysis records, if there are 

any, to determine intakes of 232Th. If a person was potentially exposed to thorium but was not 

monitored for thorium, the next paragraph provides a way to estimate intake from later whole-body 

counts.

Except for small-scope, limited-time projects that might have occurred occasionally (such as shipping 

of stored thorium from the PUREX campaigns), exposure to thorium at Hanford stopped in 1970.

Nonetheless, later whole-body counts might still be used as a check on the intakes that are estimated 

for exposure before 1970. Thorium-232 was not routinely reported in whole-body counts, but a 

prominent 228Ac peak would have been noticed and investigated. Tables 5-40 and 5-41 provide the 

ratios of 232Th to 228Ac (for types M and S, respectively) in the whole body at times after cessation of 

chronic exposure. The MDA for 228Ac or 232Th was not listed in program documentation after the 
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1960s Jech letters [1967–1969, p. 3; 1969] until 1979 when a few counts for 232Th are listed with a 

detection limit of 0.5 nCi (based on a query of the REX database). This was probably a decision level.

Twenty-one 232Th counts were recorded in the 1980s with count-specific detection limits that vary from 

0.19 to 0.93 nCi, with a median of 0.43 nCi (based on a query of the REX database). However, there 

was a difference between the detection level for a radionuclide specifically targeted for identification 

by the software and the trigger level for flagging a peak that was not specifically earmarked by the 

software. Stated another way, the criterion to identify and flag the 228Ac peak would have been higher 

than just the decision level. With that in mind, it is reasonable to assume that from 1970 through July 

1989, an 228Ac body burden of 2 nCi would have been noticed and investigated but that burdens less 

than that might have been overlooked [86]. 

Table 5-40. Ratios of 232Th to 228Ac in the whole body after end of 1- to 6-yr 

intake periods, type M [88].

Years after end 

of chronic intake

1-yr 

intake

2-yr 

intake

3-yr 

intake

4-yr 

intake

5-yr 

intake

6-yr 

intake

0.5 34.1 34.0 31.0 27.5 24.4 21.8

1 36.3 32.8 28.7 25.2 22.4 20.1

1.5 33.9 29.5 25.7 22.7 20.3 18.4

2 29.9 25.9 22.8 20.4 18.4 16.9

2.5 26.1 22.8 20.3 18.4 16.8 15.5

3 22.8 20.3 18.4 16.8 15.5 14.4

4 18.3 16.6 15.4 14.3 13.4 12.6

5 15.3 14.2 13.3 12.5 11.9 11.3

6 13.2 14.5 11.8 11.3 10.8 10.4

7 11.8 11.2 10.7 10.3 9.91 9.58

8 10.7 10.2 9.87 9.53 9.22 8.96

9 9.84 9.50 9.19 8.92 8.67 8.44

10 9.17 8.89 8.64 8.41 8.20 8.01

11 8.63 8.39 8.18 7.98 7.80 7.64

12 8.17 7.97 7.79 7.62 7.46 7.31

13 7.78 7.61 7.45 7.30 7.16 7.03

14 7.44 7.29 7.15 7.02 6.89 6.77

15 7.14 7.01 6.88 6.77 6.65 6.54

16 6.88 6.76 6.64 6.54 6.43 6.33

17 6.64 6.53 6.43 6.33 6.23 6.14

18 6.42 6.32 6.23 6.14 6.05 5.97

19 6.23 6.13 6.05 5.96 5.88 5.81

20 6.04 5.96 5.88 5.80 5.72 5.65

5.3.5 Head Counters and Other Counts

Miscellaneous counts have been performed over the years at Hanford, including wound counts, head 

counts, liver counts, lymph node counts, and various longitudinal scans with collimated detectors to 

pinpoint the location of external or internal contamination. Results of these in the database are almost 

always listed as special counts associated with known intakes [87].

Since at least 1978, for intakes of plutonium or americium, head counts have been used to correct 

chest counts for activity in the bones of the chest region [Palmer and Rieksts1979]. Since the mid1990s, liver counts have been added to the protocol for correcting chest counts to account for 

possible shine from the liver [90].

Routine head counting for 90Sr or 147Pm did occur for a while in the 1970s. The instruments used for 

these measurements were not very sensitive and there is the question about what a head count 
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Table 5-41. Ratios of 232Th to 228Ac in the whole body after end of 1- to 6-yr 

intake periods, type S [89].

Years after end 

of chronic intake

1-yr 

intake

2-yr 

intake

3-yr 

intake

4-yr 

intake

5-yr 

intake

6-yr 

intake

0.5 6.34 5.23 4.60 4.19 3.94 3.76

1 5.00 4.36 3.96 3.70 3.54 3.41

1.5 4.23 3.82 3.55 3.38 3.30 3.18

2 3.73 3.45 3.26 3.15 3.06 3.00

2.5 3.39 3.19 3.07 2.98 2.92 2.88

3 3.15 3.01 2.92 2.86 2.82 2.80

4 2.86 2.79 2.75 2.72 2.71 2.70

5 2.72 2.68 2.67 2.66 2.66 2.66

6 2.65 2.64 2.64 2.64 2.64 2.65

7 2.62 2.63 2.63 2.64 2.65 2.66

8 2.63 2.64 2.65 2.66 2.67 2.68

9 2.64 2.65 2.67 2.68 2.69 2.71

10 2.66 2.68 2.69 2.71 2.72 2.73

11 2.69 2.71 2.72 2.74 2.75 2.77

12 2.72 2.74 2.75 2.77 2.79 2.80

13 2.75 2.77 2.79 2.80 2.82 2.83

14 2.79 2.80 2.82 2.83 2.85 2.87

15 2.82 2.83 2.84 2.87 2.90 2.91

16 2.86 2.87 2.89 2.91 2.93 2.94

17 2.89 2.91 2.93 2.94 2.96 2.98

18 2.93 2.95 2.96 2.98 3.00 3.01

19 2.96 2.99 3.00 3.02 3.04 3.05

20 3.00 3.02 3.04 3.06 3.07 3.09

means in relation to the activity in the total skeleton. The same worker could have 90Sr urinalysis 

results. The latter should be given preference for confirming or quantifying an intake.

5.3.6 General Notes about Items in the Database

All in vivo results appear to be given in nCi [91]. “Limits” were MDAs, which were treated the same as 

decision levels until 1992. The decision level is listed under “limits” starting in January 1992 [92].

Sometimes a radionuclide is listed without a value or limit, which probably means a detection of what 

was considered a trace amount. More information might be available on the In Vivo Counter Results 

Form if it was sent to the worker’s personal radiation exposure history file. If not, dose reconstructors 

should assume the result of the count is 100 nCi [Glenn 1968]. 

Before the advent of GeLi detectors, when a significant peak in a whole-body count of a radionuclide 

other than 24Na, 137Cs, 40K, or 65Zn occurred, the activity of the trace or “interfering” radionuclide might

have been quantified. In addition, the activity of one or more of the regular four radionuclides might 

have been marked as invalid because of overlap with the interfering peak or because of the impact of 

the interfering peak on the spectrum stripping calculations [93]. For the small activities that are 

involved, there is no merit in trying to recalculate or estimate actual quantities. It is favorable to 

claimants to use the activities of 137Cs as given and to include the activity of the interfering 

radionuclide as given as well. A value of 100 nCi for the interfering radionuclide is appropriate if it is 

not given directly [Glenn 1968]. 

The radionuclides that have been routinely reported to the database have changed over the years.

From the beginning until 1983, 24Na, 40K, 137Cs, and 65Zn were the only routinely reported 

radionuclides. In 1983, as part of the switch to the ORE database, only 40K and 137Cs results (or the 
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MDAs) were routinely reported, and in late 1987, 60Co was added [94]. In 1995, with the 

implementation of a new spectrum analysis software program (NEXEC), the standup counter’s energy 

spectrum was divided into 12 regions and a radionuclide was assigned to each region, including more 

naturally occurring radionuclides such as 214Bi and 208Tl [95]. During this time, if a worker had a count 

from the coaxial HPGe whole-body counter, as many as 20 radionuclides might have been listed in 

the records. The listing of that many radionuclides was simply a bookkeeping approach, and had 

nothing to do with the sources of exposure [96]. Because of the shutdown of the last reactor in 1986, 

radionuclides such as 59Fe, 24Na, 22Na, 144Ce/Pr, and 131I had decayed away to negligible levels at 

Hanford (unless a researcher was using a small source for studies). The lack of the need to report all 

these radionuclides routinely, unless a peak was actually present, was recognized, and when NEXEC 

was replaced by ABACOS (October 1999), the routinely reported list was reduced to 40K, 60Co, 137Cs, 

and 154Eu [Bihl 1998] [97]. Reporting of radionuclides at levels below the MDA or decision level should 

not be interpreted as implying exposure to those radionuclides. 

For chest counting, the database usually lists 234Th as the potentially measured radionuclide as an 

indicator of 238U. Until recently, routinely reported radionuclides for anyone who received a chest 

count were 241Am, 234Th, and 235U. This does not imply exposure to both plutonium and americium 

mixtures and uranium. Very recently, workers have been scheduled for types of chest counts based 

on their exposure in the workplace, so for plutonium workers, for instance, only 241Am results are 

determined and reported.

5.4 MIXTURES 

Except for a few production facilities (such as B Plant or the Waste Encapsulation and Storage Facility 

after 1968, and the UO3 Plant), bioassay methods did not measure all the radionuclides in the intake 

mixture. The emphasis was on measuring exposure to radionuclides with the greatest impact in 

relation to radiation protection standards (for instance, MPBB or committed effective dose equivalent), 

or on radionuclides that were most common. Unmeasured radionuclides generally do not have a large 

impact on dose but might target different organs or might have a larger relative impact over periods of 

less than 50 years. Therefore, this section attempts to estimate possible mixtures of radionuclides that 

might have been part of an intake that was indicated by a measured radionuclide. In cases where

actual bioassay data are available, dose reconstructors should use those data in preference to the 

following conservative mixtures. 

Plutonium isotopic mixtures and uranium isotopic mixtures are discussed in Sections 5.2.1 and 5.2.5, 

respectively.

5.4.1 Fission and Activation Products

As presented in Section 5.1.3, an SEC was designated in which NIOSH found that it was infeasible to 

reconstruct doses for fission products before August 31, 1946, due to inadequate monitoring.

Therefore, dose reconstructions will not include any doses from fission products before August 31, 

1946.

For fission and activation product mixtures through 1988 (1 year after N Reactor shutdown), dose 

reconstructors should use the intake mixture ratios in the latest revision of ORAUT-OTIB-0054, 

Fission and Activation Product Assignment for Internal Dose-Related Gross Beta and Gross Gamma 

Analyses [ORAUT 2015]. Therefore, if an intake of 90Sr, 137Cs, or 106Ru was assigned to a worker, the 

other fission and activation products should be assigned as discussed in that document. ORAUTOTIB-0054 includes guidance on intake mixtures that applies to workers in different types of work or 

facilities.
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The fission product urinalysis procedure measured beta activity from any radionuclides of strontium, 

yttrium, barium, lanthanum, cerium, europium, and promethium. It was calibrated for the 90Sr/90Y betas 

and so would have underestimated soft beta emitters, but chemical yield was several times higher for 

cerium and yttrium than for strontium, so the recorded result overestimated cerium and yttrium activity 

[Healy 1948; Wilson 1987]. Urine results designated as FISPR or FP in the workers’ records indicate 

participation in the fission product bioassay program.

At Hanford some fission products had greater prominence than indicated in ORAUT [2015]. These 

were 65Zn for essentially all workers because of its presence in the Richland drinking water through 

about 1972 and 24Na for workers at the once-through-cooled production reactors through about 1971 

(see Section 5.3.1), 154Eu for workers at N Reactor because of the samarium-ball emergency 

shutdown system, and 24Na for 400 Area workers at the FFTF for 1980 through 1994 [98]. From 1960 

to the present, these radionuclides would have been measured by whole-body counts. If their activity 

is not included in the whole-body count record, dose reconstructors should use the standard missed 

dose procedure for assigning these intakes to workers at the applicable facilities using MDAs or 

reporting levels in Table 5-34. The N Reactor and FFTF intakes should not be applied unless it is 

certain the worker had exposure at these facilities because, relative to the Hanford workforce, few 

workers were actually exposed to these radionuclides at those times. Zinc-65 and 24Na were not 

measured as part of the fission product urinalysis, however, so during the period before whole-body 

counting, intakes of these radionuclides should be added using the guidance in Attachment C for the 

period from startup to 1959. Europium-154 is not addressed in Attachment C, and MDAs are not 

available in Table 5-34 for all years. When they are not listed in Table 5-34, dose reconstructors 

should assume the MDA is 3 times the MDA for 60Co for 1964 through 1984 [99].

ORAUT [2015] provides guidance on fission and activation product ratios for waste management 

workers. However, characterization of the contents of the high-level waste tanks shows that plutonium 

and americium are also present in sufficient ratios relative to fission products that they might increase 

the dose to certain organs. Americium is usually associated with 90Sr in the supernatant and is 

generally more available as a contaminant. Plutonium is usually associated with 137Cs in the sludge 

[Carbaugh 1995; Boothe 1992]. For workers in waste management facilities or 200 Area workers in 

general if the work location is not known, and if an intake of fission products is determined and the 

worker does not have bioassay for plutonium or americium after the intake date, dose reconstructors 

should assume an intake ratio of 0.15 for either 239Pu (type M, S, or SS) or 241Am (type M or S) 

(dependent on which is more favorable to the claimant) relative to the highest fission product intake.

The ratio is based on the 95th percentile of alpha to beta/gamma ratios currently in the high-level 

waste tanks. Current conditions have been shown to generally have higher ratios than past conditions 

[100].

Tritium and 131I intakes might have occurred in reactor workers and fuel dissolution plant workers.

Dose from tritium was accounted for as part of external dose records for 1949 through 1987.

Section 5.6 contains guidance on tritium and 131I intakes for other years or if there is no evidence of 

monitoring for tritium in the external dose record for reactor or fuel dissolution plant workers.

5.5 INTERFERENCES AND UNCERTAINTIES

5.5.1 Contamination of Samples

Home collection of excreta samples started very early in the bioassay program [Wilson 1987], so

contamination of excreta samples can be assumed to be negligible. Laboratory contamination and 

mix-up of samples in the laboratory are a possibility, although laboratory quality control procedures 

and performance of test samples were designed to minimize this source of contamination.

It is likely that a contaminated sample shows as an obvious outlier in the dataset for a given worker.
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For in vivo measurements, contamination can occur as external to the body or, in the case of chest 

counting, as external to the lung. If a followup in vivo count on the same day or within a few days 

shows a dramatic decrease in activity or no detectable activity, dose reconstructors should assume 

external contamination. Radon progeny and medical diagnostic or therapeutic procedures that involve 

radionuclides can cause interferences to in vivo measurements, especially for NaI detectors.

However, unless the count was invalidated or noted as being influenced by such interferences, the 

results should be used as recorded.

5.5.2 Uncertainties

Uncertainties for the bioassay measurements were included in the database starting in late 1981 for 

excreta measurements [101]. These are listed in the database under Error and represent the TPU 

(1 sigma) including counting uncertainty, yield uncertainty, and various other systematic uncertainties 

[UST 1989]. Dose reconstructors should use these when available. For excreta, uncertainty can also 

exist in the sample date. For routine samples, dose reconstructors should assume an uncertainty of 

±2 weeks. This is because one sample date is used for the month regardless of when the sample was 

actually obtained. For special samples an uncertainty of ±2 days is reasonable unless the sample is 

within 2 to 3 days of a known intake.

The period the sample represents is also a source of uncertainty. Most urine samples at Hanford were 

24-hour simulated samples (kit code 1), which means the sample was collected over two eveningthrough-morning periods. Medley et al. [1994] indicated that this sampling method produced only 

about half of a true 24-hour sample based on volume for a group of nine workers over a 3-day period.

However, Hanford collection protocol was based on percent of day, not volume, so the true bias 

(when samples were collected according to procedure) was about 75% of a true 24-hour sample 

[102]. If a worker has enough urine samples to establish the individual-specific excretion pattern, a 

sample can be normalized to the individual’s expected 24-hour excretion. For in vivo results, 

uncertainties were not reported until 1986 for detected radionuclides and 1993 for the default set of 

radionuclides. These were 1-sigma counting errors until 1995 [103]. The TPU has been determined 

and submitted to the records since then. The TPU includes counting uncertainty, calibration 

uncertainty, and a generic 5% positioning error (for both whole body and lung). The calibration 

uncertainty includes the uncertainty in source activity, counting error, decay correction, and 

interpolation using the calibration curve [104]. Uncertainty in reproducibly positioning a person to get 

the same result was studied at Hanford and found to be about 5% [105]. All calibrations use 

phantoms, and there is considerable uncertainty in the representativeness of phantoms versus 

humans. A recent study for whole-body counting at Hanford used a 95th-percentile Reference Man 

phantom. There was a low bias of about 25% for the coaxial HPGe detector system for 662- and 

1,332-keV gamma rays [Lynch 2007]. A similar value of uncertainty (±25%) can reasonably be 

assumed for the other whole-body detectors (1-m arc, shadow shield, and standup counters) [106].

Uncertainties in chest counting are reduced by use of different calibrations for different chest wall 

thicknesses and use of ultrasound to measure chest wall thickness. One study showed a 1-sigma 

uncertainty of about 20% for americium and uranium values in chest counting, not including correction 

for interferences from bone and liver [BMI 1990, Table 7-4]. Uncertainties would be much higher for 

an individual with activity in the bone or liver. The uncertainty in lung activity estimates that are 

affected by contributions from activity in the liver and skeleton would probably range from 100% or 

more for levels near or below the MDA to 50% or more for activity above the MDA [107]. The 

uncertainty in the estimate of chest thickness using the height-to-weight correction ratio was at least 

50% for the front to back lung counter.
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5.6 UNMONITORED WORKERS

For unmonitored workers with external dosimetry, dose reconstructors should use the intakes from the 

Attachment C co-exposure analysis in combination with the special instructions in this section and 

subsections. When assigning intakes of fission products, intakes of associated radionuclides as 

discussed in ORAUT-OTIB-0054 [ORAUT 2015] should be included. Short-lived radionuclides, such 

as 144Ce, 106Ru, 95Zr, 95Nb, 54Mn, 65Zn, 140Ba, 140La, 51Cr, 59Fe, and 110mAg, should not be assigned to 

workers after December 31, 1988 [108]. The exception is for 24Na for workers at the FFTF, which 

operated intermittently from 1980 to 1994. Special consideration is also provided for decontamination 

and decommissioning (D&D) workers. Section 5.6.2 provides facility-specific details for assignment of 

intakes and doses for unmonitored workers.

Under certain conditions, airborne effluents from one facility became air intakes for other facilities. In 

addition, workers were exposed to diluted effluents when walking between buildings or parking lots or 

while driving on the site. Therefore, workers in buildings who did not enter contaminated or airborne 

areas and construction workers almost anywhere could have incurred environmental intakes.

Before 1994, workers with even a remote chance of exposure to workplace external radiation or 

workplace airborne contamination (rather than exposure from effluents) wore a dosimeter. Through 

1971, workers obtained film dosimeters before proceeding past badge houses at the various facilities.

From 1972 to 1994, even workers with only a small potential for incurring external dose wore a 

minimal dosimeter that was called the Hanford basic dosimeter [ORAUT 2010b]. 

The unmonitored internal dose for workers with no bioassay (or just baseline or termination bioassay) 

and no evidence of ever having worn a dosimeter should be based only on environmental intakes 

[109]. Termination whole-body counts were often provided to any worker regardless of exposure risk 

and should not be considered evidence of being in a bioassay program.

The co-exposure intakes in Attachment C do not address radioiodines or tritium. The rationale for 

intake values for these radionuclides is presented in Section 5.6.1, and the assignments for dose 

reconstruction based on Section 5.6.1 and Attachment C for unmonitored workers are presented in 

Section 5.6.2. 

5.6.1 Radionuclides Not Presented in Attachment C

Guidance for assignment of intakes of radioiodines, tritium, 210Po, 241Am, 244Cm, and 99Tc by 

unmonitored workers is provided in this section.

Radioiodine

A considerable number of thyroid scans (hundreds per month) were routinely performed on workers in 

the fuel dissolution (canyon) buildings starting in 1945 and continuing into the 1950s [Parker 1948, 

p. 9; Parker 1950b, p. 8; Parker 1951, p. 8; Parker 1952, p. 9]. A review of the thyroid monitoring 

program was conducted for the Hanford Environmental Dose Reconstruction Project [Ikenberry 1991]. 

The tolerance level for 131I in air had been established in October 1945 as 1 × 10-7 μCi/cm3 [Cantril 

1945] based on a permissible equilibrium amount in the thyroid of 2 μCi. The routine scanning 

program began in late May or early June 1945. Ikenberry reviewed over 7,900 thyroid checks from 

October 1944 through August 1946. The decision level for the measurement was estimated to 

correspond to a thyroid burden of about 27 nCi, and 93% of the measurements were below this 

activity. The highest measurement that was considered to be reasonably due to a thyroid burden 

(rather than to external contamination) was 131 nCi.

Since the beginning of routine whole-body counting, detections of 131I have been rare. For instance, 

from 1962 through 1969 (the peak production years), only about 130 of about 14,500 workers who 
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had whole-body and thyroid counts had detectable 131I. The results ranged from 0.001 to 560 nCi, and 

the highest was a same-day measurement after an accidental exposure in the T Plant canyon. Ninety 

percent of the detected results were less than 15 nCi [110]. From these two datasets, the assumption 

was made that was favorable to claimants that at-risk workers had thyroid burdens less than 30 nCi, 

which would result from a daily intake of 20,000 pCi [111]. This intake would apply to all at-risk 

workers before 1960. From 1960 to 1987, it is reasonable to assume that an unmonitored worker (i.e., 

no whole-body count) did not incur intakes greater than the thousands of workers who had wholebody counts that showed no detection. Using a nominal detection limit for whole-body counts of 5 nCi 

(rather than more sensitive thyroid counts), unmonitored workers that are associated with reactors or 

fuel dissolution plants should be assigned 131I intakes of 3,400 pCi/d [112]. No intakes of 131I should be 

assigned after 1987 because there would have been no significant source term unless the worker’s 

personal information indicates a special exposure, such as for a research project [113].

Tritium

Tritium bioassay results were not included in the database until 1984 [114]. Tritium doses were 

included in external dose reports from 1949 through 1987, so monitored workers would be determined 

by tritium dose in external dose records. A zero dose is indicative of a monitored worker, but there are 

only two of these in the database. No value listed at all for tritium might be indicative of either an 

unmonitored worker or a monitored worker with a less-than-recordable dose. No tritium doses are 

listed on external dose reports for 1955 through 1960, 1970, 1971, and 1973 through 1984, so either 

there was no bioassay monitoring during these periods or the intakes were below some reporting 

criteria [115]. The 108-B Tritium Extraction Facility shut down in 1955, which probably explains the 

lack of tritium doses, and probably a lack of tritium monitoring, until the heavy-water-moderated PRTR 

started up in 1960. If a worker’s file does not show tritium dose in the external dose records (where a 

zero is considered a monitored dose), and if the worker might have had exposure to tritium as 

discussed below, then the unmonitored tritium intakes discussed below and provided in Table 5-42 

should be applied.

Table 5-42. Statistical parameters of Hanford recorded tritium doses and associated intakes [116].a

Period

GM 

(mrem) GSD

GM doses normalized 

to present model 

(mrem)

Daily intakes resulting 

in the normalized 

doses ( × 106 pCi/d)b

1949 29.9 2.9 (3.0)c 48.4 2.0

1950 43.6 2.8 (3.0) 70.6 2.9

1951 64.2 3.5 104 4.2

1952 67.1 4.0 109 4.4

1953 40.8 3.7 38.5 1.6

1954 104 3.0 98.2 4.0

1955 40.8 3.7 38.5 1.6

1956–1960 Insufficient data Insufficient data Insufficient data Insufficient data

1961 62.4 1.9 (3.0) 58.9 2.4

1962 193 3.0 182 7.4

1963 185 3.8 175 7.1

1964 141 4.2 133 5.4

1965 134 3.8 127 5.1

1966 108 3.9 102 4.1

1967 185 4.7 175 7.1

1968 98.6 4.9 93.1 3.8

1969 26.5 2.3 (3.0) 25.0 1.0

1970–1971 Insufficient data Insufficient data Insufficient data Insufficient data

1972 58.9 2.3 (3.0) 55.6 2.3

a. GM = geometric mean; GSD = geometric standard deviation.

b. Use as injection mode in IMBA or Chronic Annual Dose Workbook.

c. A default GSD of 3.0 should be substituted for a calculated GSD of less than 3.0.
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Before 1949, it is assumed that tritium intakes were limited by the tolerance level. The earliest tritium 

tolerance limit that has been found to date is an air concentration limit of 0.01 μCi/L in 1947 [Parker 

1947; Patterson 1949] (see Attachment A for discussion of tolerance levels). The assumption of a 

chronic intake at this concentration of 1.2 m3/hr for 2,000 hr/yr results in an intake rate of 

6.6 × 107 pCi/d. This produces annual organ doses of about 1,500 mrem [117], which is larger than all 

but a few recorded doses from 1949 through the 1960s. Therefore, this intake rate is favorable to 

claimants for the highest risk workers, namely those at the 108-B Tritium Extraction Facility. Tritium 

would have been present at the fuel dissolution plants as well and to a lesser extent at the reactors, 

but the tritium in the latter facilities would certainly not have been as concentrated as at the 108-B 

facility. It is assumed that tritium intakes at the reactors and fuel dissolution facilities would have been 

at most 10% of the intakes at 108-B [118]. Therefore, daily intakes of 7 × 106 pCi/d should be 

assigned to unmonitored workers at the reactors, fuel dissolution plants, or rovers.

To determine tritium intakes for 1949 through 1972, a co-exposure analysis was conducted on tritium 

dose data for that period. The statistical results are biased high because, with two exceptions, doses 

from tritium of <10 mrem were not recorded; consequently, it is not known how many workers were 

monitored who had doses less than 10 mrem. The annual geometric means (GMs) and geometric 

standard deviations (GSDs) are shown in Table 5-42.

The methodology for calculating tritium dose from urine concentration has changed over time. The 

method introduced in ICRP Publications 2 and 10 [ICRP 1959, 1968] resulted in dose per unit activity 

in urine slightly higher than the method used in the present. How the tritium doses were calculated 

before 1959 has not been found yet, but Parker, the head manager of radiation protection at Hanford 

in the 1940s and 1950s, referenced a method by Morgan in 1946 that shows that 40,000 μCi in the 

body produces the tolerance dose for the time [Morgan 1947]. The tolerance dose in the 1940s was 

100 mrep/d. Under the assumption that the tritium is distributed in 42 L of body water at equilibrium, 

952 μCi/L (40,000 μCi/42L) produced 100 mrem/d. Stated another way, a dose of 0.105 mrem/d 

would have been calculated from 1 μCi/L in urine. The 1953 body burden of tritium to produce a 

maximum permissible dose of 300 mrem/wk was 10,000 μCi [NBS 1953]. Therefore, 238 μCi/L 

(10,000 μCi/42L) would have produced 42 mrem/d, or 0.18 mrem/d would have been calculated from 

1 μCi/L in urine. In the present model [ICRP 1997], 1.2 × 106 Bq/L results in 20 mSv/yr, or a 

0.17 mrem/d dose is calculated from 1 μCi/L in urine. Therefore, the recorded doses for 1953 through 

1972 are slightly higher per unit activity in urine than would be calculated under the present model.

Assuming Hanford used the Morgan model through 1952, those recorded doses were increased by 

0.17/0.105 to be consistent with the present model.

After the above calculations were made, hundreds of tritium bioassay records were found in archived 

records and analyzed. The results are discussed in Attachment C. The analysis of actual tritium 

bioassay records showed that the above recommendations based on tritium dose records were 

favorable to claimants.

Because the doses in Table 5-42 were based on only the nonzero recorded doses, they were judged 

to be most applicable to unmonitored workers at the 108-B facility through 1955 and the PRTR from 

1960 through 1969. The 108-B Tritium Extraction Facility shut down in 1955, and some tritium 

bioassay has been found for that year. Those records show that the 1953 default doses and intakes 

are favorable to claimants if they are also used for 1955. It is assumed that tritium doses at the 

reactors and fuel dissolution facilities would have been at most 10% of the doses in Table 5-42 [119].

The GSDs in Table 5-42 should also be used with a minimum GSD of 3.

There was insufficient dose data in two periods, 1956 to 1960 and 1970 to 1971, for statistical 

analysis. Disregarding improvements in radiation protection regulations and practices, tritium 

exposure at the reactors and fuel dissolution plants could have been correlated with production. Using 

tons of uranium times fuel burnup as a figure of merit for production (see Figure C-1 in Attachment C), 
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average yearly production in 1955 through 1960 was 4 times greater than for 1949 through 1954. The 

average of the median daily intakes for unmonitored reactor and fuel dissolution plant workers during 

1949 through 1954 was 3.2 × 105 pCi/d (one-tenth of the average of the intakes in the fifth column in 

Table 5-42). Four times this intake, 1.3 × 106 pCi/d, should be used for unmonitored reactor and fuel 

dissolution plant workers during 1956 through 1960. The largest GSD of 4 can be applied to the 1956 

through 1960 intakes. Production was essentially equal during 1970 and 1971 in comparison to 1949 

through 1954, so the 3.2 × 105 pCi/d intake should be applied to unmonitored reactor and fuel 

dissolution plant workers during 1970 and 1971 (GSD = 4).

No fuel dissolution occurred from 1973 through 1983 and only one reactor was operating. The 

3.2 × 105 pCi/d intake can also be applied to unmonitored reactor workers during this period. A small 

set of urinalyses were collected from reactor workers in 1982 to 1983. The set was too small for a 

complete statistical analysis, but the median result was about 4 pCi/mL. This results from daily intakes 

of about 1.2 × 104 pCi/d [120] and indicates that the 3.2 × 105 pCi/d intake rate is favorable to 

claimants.

Tritium bioassay results are available for 1984 through 1986 for 100 and 200 Area workers. A coexposure analysis was performed on these data [121]. The 50th-percentile urinalysis results were 2.6, 

0.37, and 0.21 pCi/mL for 1984, 1985, and 1986, with GSDs of 13, 3.3, and 6.2, respectively. The 

associated daily intakes were 7,600, 1,080, and 610 pCi/d, respectively [122]. N Reactor shut down in 

1987; PUREX ran intermittently in the late 1980s and was shut down in 1992 (see Section 2.2.3 of the 

Hanford Site Description, ORAUT [2010a]). The 1986 value can be applied to workers at those 

facilities through their respective shutdown dates [123].

By 1992 there was little exposure to tritium at Hanford except for special projects (e.g., tritium target 

program), low-level tracers, or in 400 Area drinking water as discussed below.

For workers in the Hanford 400 Area, of which the principal facility was the FFTF, drinking water was 

obtained from an unconfined aquifer that had low-level tritium contamination. The FFTF was started in 

1980. The exact date well water was provided for drinking at the site has not been discovered, but it is 

reasonable to assume that it predated the startup of the FFTF by a year or so. An analysis of the 

tritium concentrations in the wells that provide the drinking water and estimation of the tritium intakes 

from that source was performed [ORAUT 2005a]. Some 400 Area workers were on routine tritium 

bioassay, and dose reconstructors should use those data when available. For other 400 Area 

workers, calendar day ingestion intakes of 48,000 pCi/d for 1978 through 1985 and 14,000 pCi/d for

1986 through present should be assigned [ORAUT 2005a]. However, the annual doses from the post1985 intakes are less than 1 mrem and can be ignored [124].

Unmonitored intakes for certain radionuclides are determined differently. Generally these were 

radionuclides not related to the primary radionuclides or mixtures (i.e., plutonium, uranium, tritium, 

and mixed fission and activation products). Bioassay monitoring for the secondary radionuclides at 

certain times was rare or nonexistent. Most of the exposure to these other radionuclides is covered 

under the SEC.

In general, workplace monitoring and action levels, such as for air concentrations, were based on total 

alpha or total beta measurements with action levels set for the most restrictive radionuclide of the list 

of radionuclides likely to be encountered. For instance, the total alpha measurements were usually 

compared to the MPC for 239Pu and the total beta measurements were compared to the MPC for 90Sr. 

An exception was for air samples collected at facilities where uranium was the predominant 

radioelement. Containments and other barriers (e.g., gloveboxes, hot cells) were also similar for 

radionuclides emitting similar radiations. Protocols for establishing of contamination zones and 

performing workplace and person surveys and other radiation protection practices were the same for 

the secondary radionuclides as for the primary radionuclides. Therefore, it is reasonable to apply the 
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plutonium co-exposure intakes (Attachment C) to unmonitored intakes of 241Am, curium, 210Po, and 

237Np. Thorium intakes should be based on co-exposure intakes for uranium. However, these 

guidelines do not apply to all times. In addition, there are situations when the 95th percentile of the coexposure distribution and a constant distribution are more appropriate than the 50th percentile and 

lognormal GSDs. The 95th percentile is used when routine bioassay was not available or not used.

The 95th-percentile intake can be found using:

(5-1) 1 645 - - . 95th percentile intake 50th percentile intake = × GSD

Specifics are given for each situation below.

Polonium-210

Exposure to 210Po before July 1, 1972, is covered by the SEC because it was determined to be 

infeasible to reconstruct doses. Bioassay monitoring was used for significantly exposed workers from 

1968 on. If an intake estimate is required, unmonitored workers from July 1, 1972, on should be 

assigned intakes at the same values as the plutonium co-exposure intakes using the 50th-percentile 

values and the GSDs from Attachment C.

Americium-241

Significant quantities of pure 241Am or mixtures enriched in 241Am were present in several Z Plant 

facilities through 1976 and at times at a few other facilities. See the Site Description [ORAUT 2010a]

for details. Exposure before July 1, 1972, is covered by the SEC because it was determined to be 

infeasible to reconstruct doses. Bioassay monitoring via chest counts was used for significantly 

exposed workers from 1968 on, and urinalysis was used for special monitoring after incidents with 

potential for intake. If an intake estimate is required from July 1, 1972, on, unmonitored workers 

should be assigned intakes at the same values as the plutonium co-exposure intakes using the 50thpercentile values and the GSDs from Attachment C.

Curium-244

Curium-244 was extracted and purified from Shippingport fuel in 1966 through 1968 at the 

201-C facility and the hot cells in the 325 Building. Unrelated work with curium was performed in the 

308 Building from 1969 through 1973. See the Site Description [ORAUT 2010a] for details. Exposure 

before July 1, 1972, is during the SEC, and intakes will not be assigned for this period for partial dose 

reconstructions for unmonitored workers. Incident followup urinalysis for curium was available and 

used rarely from 1968 on. Limited use of routine urinalysis was conducted starting in 1974. If an 

intake estimate is required, unmonitored workers should be assigned intakes at the 95th percentile of 

the plutonium co-exposure intakes (constant distribution) for before 1974 and at the 50th percentile 

(with GSDs) for 1974 on.

Technetium-99

Technetium-99 was extracted and purified from fuel in a few campaigns during 1962 through 1964 

with later shipments of the material in 1967 and 1970. The work was mostly conducted in the hot cells 

in the 325 Building, although some or all of the purified material was subsequently stored somewhere 

in the 200 Area. The 1970 shipment involved a mix of previously extracted material with recently 

extracted material with the work being performed at the REDOX Plant. No evidence of significant 

exposure to 99Tc after 1970 has been found. No evidence of bioassay for 99Tc in the 1960s or 1970s 

has been found, so the exposure is considered unmonitored. Exposure before July 1, 1972, is during

the SEC, and intakes will not be assigned for this period for partial dose reconstructions for 

unmonitored workers.
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5.6.2 Assignments for Unmonitored Workers

As presented in Section 5.1.3, an SEC was designated in which NIOSH found that it was infeasible to 

reconstruct doses for plutonium or fission products before August 31, 1946, due to inadequate 

monitoring. Therefore, dose reconstructions will not include any doses from plutonium or fission 

products before August 31, 1946. The first extruded uranium rods arrived in October 1943 and 

machining began in December.

In an addition to the SEC, NIOSH found that it was infeasible to reconstruct doses for employees of 

DOE contractors and subcontractors, except for those of employers excluded (see Section 5.1.3), 

during the period from January 1, 1984, through December 31, 1990, due to inadequate bioassay 

monitoring (Section 5.1.3). The co-exposure intake rates in Attachment C are based on bioassay data 

that do not adequately represent the added SEC class for the period from January 1, 1984, through 

December 31, 1990. Therefore, co-exposure intakes in Attachment C will not be assigned for January 

1, 1984, through December 31, 1990, for partial dose reconstructions for claims not covered within the 

SEC. 

When information about what facility or type of facility an employee worked at is available, guidance 

on how to assign mixtures of the various radionuclides is given in the facility and/or area descriptions 

following Table 5-43. Dose reconstructors should check the Site Description for years of operation of 

specific facilities [ORAUT 2010a]. It is important to note that ingestion intake of 65Zn should be 

assigned to all workers from August 31, 1946, through 1972 in accordance with Attachment C. For 

1949 to 1955, 1961 to 1969, and 1972, note that it is the tritium doses in Table 5-43 that are to be 

assigned by dose reconstructors and, if the derived daily intake values are used in a dose calculation 

tool, the calculated doses should be verified with the doses in Table 5-43. 

Table 5-43. Intakes for workers with external monitoring but essentially no bioassay (pCi/d) [125]. 

Radioactive 

material Period 

Daily intake (pCi/d), 

dose from tritium 

only (mrem/yr) Distribution Note

Pu 09/01/1946–

present

See Attachment Ca See Attachment C None

U Through present See Attachment Ca See Attachment C None

Sr-90 09/01/1946–

present

See Attachment Ca See Attachment C None

Pm-147 See Attachment 

C

See Attachment Ca See Attachment C None

Zn-65 09/01/1946–

present

See Attachment Ca See Attachment C None

Na-24 09/01/1946–

present

See Attachment Ca See Attachment C None

Cs-137 09/01/1946–

present

See Attachment Ca See Attachment C None

Other fission and 

activation 

products

09/01/1946–

present

See ORAUT-OTIB0054 [ORAUT 2015] See ORAUT-OTIB0054 [ORAUT 2015] None

I-131 (vapor) 09/01/1946–1959 2E+04 Lognormal, GSD = 3 None

I-131 (vapor) 1960–1987 3.4E+03 Lognormal, GSD = 3 N Reactor only 1973 

through 1983.

1-131 (vapor) 1988–present 0 Not applicable None

Tritium (HTO) 09/01/1946–1948 6.6E+06, 163 Lognormal, GSD = 3 Reactors, fuel 

dissolution, rovers

Tritium (HTO) 1949 2.0E+06, 48.4 Lognormal, GSD = 3 108B

Tritium (HTO) 1950 2.9E+06, 70.6 Lognormal, GSD = 3 108B

Tritium (HTO) 1951 4.2E+06, 104 Lognormal, GSD = 3.5 108B
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 73 of 180

Radioactive 

material Period 

Daily intake (pCi/d), 

dose from tritium 

only (mrem/yr) Distribution Note

Tritium (HTO) 1952 4.4E+06, 109 Lognormal, GSD = 4.0 108B

Tritium (HTO) 1953 1.6E+06, 38.5 Lognormal, GSD = 3.7 108B

Tritium (HTO) 1954 4.0E+06, 98.2 Lognormal, GSD = 3.0 108B

Tritium (HTO) 1955 1.6E+06, 38.5 Lognormal, GSD = 3.7 108B

Tritium (HTO) 1949 2.0E+05, 4.8 Lognormal, GSD = 3.0 Reactors, fuel 

dissolution, rovers

Tritium (HTO) 1950 2.9E+05, 7.1 Lognormal, GSD = 3.0 Reactors, fuel 

dissolution, rovers

Tritium (HTO) 1951 4.2E+05, 10.4 Lognormal, GSD = 3.5 Reactors, fuel 

dissolution, rovers

Tritium (HTO) 1952 4.4E+05, 10.9 Lognormal, GSD = 4.0 Reactors, fuel 

dissolution, rovers

Tritium (HTO) 1953 1.6E+05, 3.9 Lognormal, GSD = 3.7 Reactors, fuel 

dissolution, rovers

Tritium (HTO) 1954 4.0E+05, 9.8 Lognormal, GSD = 3.0 Reactors, fuel 

dissolution, rovers

Tritium (HTO) 1955 1.6E+05, 3.9 Lognormal, GSD = 3.7 Reactors, fuel 

dissolution, rovers

Tritium (HTO) 1956–1960 1.3E+06, 32 Lognormal, GSD = 4 Reactors, fuel 

dissolution, rovers

Tritium (HTO) 1961 2.4E+05, 5.9 Lognormal, GSD = 3.0 Reactors (except 

PRTR), fuel dissolution, 

rovers

Tritium (HTO) 1962 7.4E+05, 18.2 Lognormal, GSD = 3.0 Reactors (except 

PRTR), fuel dissolution, 

rovers

Tritium (HTO) 1963 7.1E+05, 17.5 Lognormal, GSD = 3.8 Reactors (except 

PRTR), fuel dissolution, 

rovers

Tritium (HTO) 1964 5.4E+05, 13.3 Lognormal, GSD = 4.2 Reactors (except 

PRTR), fuel dissolution, 

rovers

Tritium (HTO) 1965 5.1E+05, 12.7 Lognormal, GSD = 3.8 Reactors (except 

PRTR), fuel dissolution, 

rovers

Tritium (HTO) 1966 4.1E+05, 10.2 Lognormal, GSD = 3.9 Reactors (except 

PRTR), fuel dissolution, 

rovers

Tritium (HTO) 1967 7.1E+05, 17.5 Lognormal, GSD = 4.7 Reactors (except 

PRTR), fuel dissolution, 

rovers

Tritium (HTO) 1968 3.8E+05, 9.3 Lognormal, GSD = 4.9 Reactors (except 

PRTR), fuel dissolution, 

rovers

Tritium (HTO) 1969 1.0E+05, 2.5 Lognormal, GSD = 3.0 Reactors (except 

PRTR), fuel dissolution, 

rovers

Tritium (HTO) 1960 2.4E+06, 58.9 Lognormal, GSD = 3.0 PRTR

Tritium (HTO) 1961 2.4E+06, 58.9 Lognormal, GSD = 3.0 PRTR

Tritium (HTO) 1962 7.4E+06, 182 Lognormal, GSD = 3.0 PRTR

Tritium (HTO) 1963 7.1E+06, 175 Lognormal, GSD = 3.8 PRTR

Tritium (HTO) 1964 5.4E+06, 133 Lognormal, GSD = 4.2 PRTR

Tritium (HTO) 1965 5.1E+06, 127 Lognormal, GSD = 3.8 PRTR

Tritium (HTO) 1966 4.1E+06, 102 Lognormal, GSD = 3.9 PRTR
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 74 of 180

 

Radioactive 

material Period 

Daily intake (pCi/d), 

dose from tritium 

only (mrem/yr) Distribution Note

Tritium (HTO) 1967 7.1E+06, 175 Lognormal, GSD = 4.7 PRTR

Tritium (HTO) 1968 3.8E+06, 93.1 Lognormal, GSD = 4.9 PRTR

Tritium (HTO) 1969 1.0E+06, 25.0 Lognormal, GSD = 3.0 PRTR

Tritium (HTO) 1970–1971 3.2E+05, 7.9 Lognormal, GSD = 4 Reactors, fuel 

dissolution

Tritium (HTO) 1972 2.3E+06, 55.6 Lognormal, GSD = 3 N Reactor, PUREX 

plant

Tritium (HTO) 1973–1983 3.2E+05, 7.9 Lognormal, GSD = 4 N Reactor

Tritium (HTO) 1984 7.6E+03, 0.2 Lognormal, GSD = 13 N Reactor, PUREX

Tritium (HTO) 1985 1.1E+03, 0.03 Lognormal, GSD = 3.3 N Reactor, PUREX

Tritium (HTO) 1986–1987 6.1E+02, 0.02 Lognormal, GSD = 6.2 N Reactor, PUREX 

Tritium (HTO) 1988–1992 6.1E+02, 0.02 Lognormal, GSD = 6.2 PUREX only

Tritium (HTO) 1978–1985 4.8E+04, 1.2 Lognormal, GSD = 3 Ingestion of drinking 

water in the 400 Area

Am-241 (type M) 07/01/1972–

present

Use Pu intakes (type 

M) from Attachment C

50th percentiles with 

GSDs

None

Cm-244 (type M) 07/01/1972–

12/31/1973

Use Pu intakes (type 

M) from Attachment C

95th percentiles, 

constant distribution

None

Cm-244 (type M) 1974–present Use Pu intakes (type 

M) from Attachment C

50th percentiles with 

GSDs

None

Po-210 (type F or 

M)

07/01/1972–

present

Use Pu intakes (type 

M) from Attachment C

50th percentiles with 

GSDs

None

Tc-99 (type F or 

M)

07/01/1972–

present

None None No known exposures 

during this period. 

D&D radiological 

workers

2002–present See discussion, last 

paragraph in this 

section. 

Not applicable None

a. Per Section 5.1.3, the co-exposure intake rates for unmonitored or inadequately monitored workers in Section 5.6 and 

Attachment C are based on bioassay data that do not adequately represent workers included in the 1984-to-1990 SEC 

class and will not be assigned for that period.

It is important to note that as presented in Attachment C, from December 1943 through 1948, intakes 

for uranium should be assigned in accordance with the Process/Job Title listed as Machining/Operator 

in the 300 Fuel Fabrication area as presented in Battelle-TBD-6000 [BT 2011].

Attachment C intake rates exhibit a decreasing trend over time and may be assigned past 1988 when 

overestimates are appropriate for dose reconstructions.

100 Area Reactors

• While operating, 1944 to 1987: Use fission and activation products (from September 1, 1946, 

on) from Attachment C (except 147Pm) and ORAUT-OTIB-0054 [ORAUT 2015] and 131I and 

tritium intakes from Table 5-43. 

• When not operating, 1988 to the present: Use fission and activation products from 

Attachment C only without 131I or tritium and without short-lived fission and activation products 

after 1988 for best estimates.

300 Area Test Reactors, 1944 to 1978

Use fission and activation products (from September 1, 1946, on) from Attachment C (except 147Pm) 

and ORAUT-OTIB-0054 [ORAUT 2015] and tritium for PRTR from Table 5-43. 
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 75 of 180

400 Area (includes 405 Building FFTF), 1980 to 1994 (1978 to 1985 for Tritium)

Use fission and activation products (reactors) from Attachment C (except 147Pm) and ORAUT [2015].

Note that this reactor continued to be used intermittently until 1994. Use tritium ingestion from drinking 

water from Table 5-43.

200 Area Fuel Separation Facilities

• T-Plant, 224-T, 1945 to 1956; B-Plant, 224-B, 1945 to 1952; REDOX, 224-S, 1952 to 1967; 

PUREX 1956 to 1992: Use fission and activation products (from September 1, 1946, on), 

plutonium (from September 1, 1946, on, types M, S, or SS), and uranium (type F from January 

1, 1949, on) from Attachment C (except 147Pm) and ORAUT [2015] but may ignore 24Na from 

Attachment C for best estimates. Use tritium and 131I intakes from Table 5-43. No short-lived 

fission and activation products should be applied after 1988 for best estimates.

• B-Plant, 225-B, 1953 to the present: Use fission and activation products from Attachment C 

(except 147Pm) and ORAUT [2015], waste management option) but may ignore 24Na for best 

estimates. Use 65Zn ingestion-only intakes for 1953 to 1972.

PFP facilities (various Z buildings and 233-S), September 1, 1946, to Present

Use plutonium intakes from Attachment C (M, S, or SS). If it is more favorable to the claimant:

• For July 1, 1972, through 1976, intakes of 241Am may be assigned in lieu of plutonium.

• For 231-Z, 241Am may be assigned in lieu of plutonium from July 1, 1972, on.

221-U, 224-U (UO3 Plant), 1952 to 1993

Use uranium type F intakes from Attachment C.

222-S Laboratory

Use fission and activation products (fuel dissolution) from Attachment C and ORAUT [2015]. Include 

147Pm for 1964 through 1965.

108-B Tritium Extraction Facility, 1949 to 1956

Use tritium intakes from Table 5-43. Exposure to organically bound tritium (OBT) and type F tritides 

were possible.

C-Plant (Hot Semiworks, Strontium Semiworks)

• 1949 to 1956: Use fission and activation products (fuel dissolution) from Attachment C (except 

147Pm) and ORAUT [2015]; 24Na may be ignored for best estimates. Use plutonium (M, S, SS) 

and uranium (F) from Attachment C.

• 1962 to 1967: Use fission and activation products from Attachment C and ORAUT [2015] (fuel 

dissolution); 24Na may be ignored for best estimates. Use 65Zn ingestion intakes only. Include 

147Pm for 1964 through 1967.

300 Area Fuel Fabrication Facilities (303, 305, 306, 313, 314, 333), 1944 to 1988

Use insoluble uranium intakes from Attachment C. It is important to note that in Attachment C, from 

December 1943 through 1948, intakes for uranium are assigned in accordance with the Process/Job 

Title listed as Machining/Operator as presented in Battelle-TBD-6000 [BT 2011].

209-E and 120 Critical Mass Laboratories, 1950 to 1986

Use plutonium intakes (M, S, or SS) from Attachment C.
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308 Plutonium Fuels Pilot Plant, 1960 to 1990

Use fission and activation products (fuel dissolution), plutonium (M, S, or SS), uranium (M or S), and 

147Pm intakes from Attachment C and ORAUT [2015]. Use 131I intakes from Table 5-43; 24Na intakes 

may be ignored for best estimates. Intakes of 244Cm from July 1, 1972, through 1973 may be used in 

lieu of plutonium if it is more favorable to the claimant.

324 Building, 1966 to 2001

Use fission and activation products (fuel dissolution) from Attachment C (except 147Pm) and ORAUT 

[2015]. The fuel was well-aged before arriving, so 24Na and other short-lived radionuclides in ORAUT 

[2015] may be ignored for best estimates and may use 65Zn ingestion intakes only. From July 1, 1972, 

on, include intakes of 241Am at plutonium co-exposure intake levels from Attachment C.

325 Building, 1953 to the Present

Use fission and activation products (fuel dissolution), plutonium (M, S, or SS), and 147Pm intakes from 

Attachment C and ORAUT [2015]. Sodium-24 may be ignored for best estimates but, because of 

multiple uses of this facility, short-lived radionuclides might have been present past 1987.

327 Building, 1953 to 1987

Use fission and activation products (fuel dissolution) and plutonium (M, S, or SS) intakes from 

Attachment C and ORAUT [2015]. Include 147Pm for 1966 through 1970; 24Na may be ignored for best 

estimates.

3706 Building (Radiochemistry Laboratory, Technical Building), 1945 to about 1963

Assign fission and activation products (from September 1, 1946, on) and plutonium (from 

September 1, 1946, on, types M, S, or SS) intakes from Attachment C (except 147Pm) and ORAUT 

[2015]. Assign uranium intakes (types M or S from January 1, 1949, on) from Attachment C.

3708 Building

Assign intakes of 241Am or 244Cm based on plutonium co-exposure intakes from Attachment C for July 

1, 1972, to 1973.

Tank Farms, Evaporators, and 340 Liquid Waste Handling Buildings, September 1, 1946 to the 

Present

Assign fission and activation products (waste management) from Attachment C (except 147Pm) and 

ORAUT-OTIB-0054 [ORAUT 2015]. Assign plutonium (M, S, or SS) or americium intakes at 

0.15 times the 137Cs fission product intake from Attachment C (to account for an alpha component in 

the waste). Sodium-24 may be ignored for best estimates, and 24Na, 65Zn, and radionuclides with 

short half-lives may be ignored after 1988.

D&D and Remediation, 2002 to the Present

For employees of the site’s remediation contractor whose job description indicates potential for 

exposure to airborne contamination, assign 40 DAC-hr of unmonitored intakes per year (based on the 

investigation level of the lapel air sampling program). Various radionuclides are possible, but choose 

the most favorable to the claimant of the following (choose one): 260 pCi/d of either 90Sr (type F) 137Cs 

(type F), 152Eu (type M), or 14C (particulate type M or S), or 0.26 pCi/d of 239Pu (type M, S, or SS) 

[126].

Rovers

Could have been exposed at many different facilities. Use the same intakes as for the fuel separations 

facilities.
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Unknown Facility

Use judgment based on any available information and ORAUT [2015]. If the work can be associated 

with the 100 Area, assume exposure at a reactor. If it can be associated with the 200 Area, then 

consider assuming the worst case, which would be the fuel separations facilities. If it can be 

associated with the 300 Area, it would be helpful to determine if the work was associated with fuel 

fabrication or one of the research facilities. After 1964, fuel fabrication was done by the same 

contractor(s) as operated the 100 Area reactors (e.g., Douglas or United Nuclear Corporation).

Assume 308 Building if no other 300 Area information is available. If the work location is listed as the 

600 Area, assign environmental intake only.

5.7 ATTRIBUTIONS AND ANNOTATIONS

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

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

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

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

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

for review on the Project’s Site Research Database (SRDB).

Donald Bihl served as the initial Subject Expert for this document. Mr. Bihl was previously employed 

at Hanford and his work involved management, direction or implementation of radiation protection 

and/or health physics program policies, procedures or practices related to atomic weapons activities 

at the site. This revision (and earlier revisions) have been overseen by a Document Owner who is fully 

responsible for the content, including all findings and conclusions. In all cases where such information 

or prior studies or writings are included or relied upon by Mr. Bihl, those materials are fully attributed 

to the source.

Fred Duncan assumed responsibility as Document Owner for this document in September, 2008.

Mr. Duncan replaced Edward Scalsky when Mr. Scalsky’s employer declared a new corporate conflict 

of interest for the Hanford Site.

[1] Bihl, Donald E. Oak Ridge Associated Universities (ORAU) Team. Principal Health Physicist.

October 2003.

This is based on personal experience. These were the protocols for reporting results from the 

bioassay laboratory to the dosimetry database.

[2] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience. These were the protocols for reporting results from the 

bioassay laboratory to the dosimetry database.

[3] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on discussions with former in vivo counting program staff [name redacted] and 

[name redacted]. The GOK region also shows on whole-body count forms from the 1960s and 

1970s.

[4] Bihl, Donald E. ORAU Team. Principal Health Physicist.

The dates for the HIE and ORE databases are from Wilson [1987]. The other remarks are from 

personal experience and discussions with various program managers of the Radiological 

Records Program.

[5] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience and review of the bioassay database.
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 78 of 180

 

[6] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience and review of the bioassay database.

[7] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

Information in the previous paragraph came from personal experience, common knowledge 

passed down from previous managers of the Hanford Internal Dosimetry Program, and review 

of the various statements of work with the bioassay laboratory.

[8] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is due to the nature of the analysis, which involved separation of the plutonium then gross 

alpha analysis. Plutonium-241 is a beta emitter, and 241Am would have been removed during 

the chemistry.

[9] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience. In addition, most of this information is readily noticeable 

based on queries of the database, which show the nature of how the data were recorded and 

when the changes in recording practices occurred. Detection levels are also discussed in 

Wilson [1987]. The recording of 0.025 dpm/sample for nondetections during 1975 to 1982 and 

recording of actual results for measured results >0.025 dpm/sample was also verified when 

the former [title redacted], [name redacted], was asked that direct question shortly after his 

retirement in the late 1980s.

[10] Bihl, Donald E. ORAU Team. Principal Health Physicist. November 2004.

This information is based on discussion in Wilson [1987] and general understanding of how 

the concepts of detection level and MDA have changed over the years.

[11] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience. In addition, most of this information is readily noticeable 

based on queries of the database, which show the nature of how the data were recorded and 

when the changes in recording practices occurred. Detection levels are also discussed in 

Wilson [1987]. The recording of 0.025 dpm/sample for nondetections during 1975 to 1982 and 

recording of actual results for measured results >0.025 dpm/sample was also verified when 

the former [title redacted], [name redacted], was asked that direct question shortly after his 

retirement, in the late 1980s.

[12] Bihl, Donald E. ORAU Team. Principal Health Physicist. November 2006.

This information was provided by [name redacted], the [title redacted], and [name redacted], 

[title redacted] in the Hanford Internal Dosimetry Program. It is also discussed in a letter to the 

Hanford Internal Dosimetry Program file [Antonio 2002]. 

[13] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience.

[14] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience and discussions with [name redacted], the [title 

redacted].

[15] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience. In addition, a summary of bioassay procedures is 

contained in BMI [2003b]. That two results were occasionally recorded for the same sample, 

because both splits were analyzed, is also evident from a query of the database.
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 79 of 180

 

[16] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

The MDAs were extracted from various statements of work for the bioassay laboratory. The 

true MDAs for various times were determined as part of the laboratory’s quality control 

program and from a double-blind oversight program operated by the Hanford Internal 

Dosimetry Program.

[17] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

The requirement to report TPU shows in the 1983 statement of work [Simpson 1983]. A query 

of the database shows that reporting of the uncertainty was sporadic in 1980 and became 

routine in 1981.

[18] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was extracted from various statements of work with the bioassay laboratory, 

including AEC [1964], Heid [1974], DOE [1979], Simpson [1983], DOE [1985, 1987], and 

Acker [1992]. 

[19] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was extracted from various statements of work with the bioassay laboratory, 

including AEC [1964], Heid [1974], DOE [1979], Simpson [1983], DOE [1985, 1987], and 

Acker [1992]. 

[20] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

There is some mention of the 238Pu operations in Gerber [1992], although that reference does 

not give the date of 1967. The original reference (an incident write-up) using the year 1967 has 

not been located.

[21] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on discussions with PUREX engineers in the 1980s and from Sula et al. [1989]. 

[22] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

The composition of the plutonium used in the Nuclear Waste Vitrification Project was added to 

the Internal Dosimetry Program TBD based on a memorandum from D. Bihl to Program Staff, 

“Unusual Plutonium Mixture” [Bihl 1992], which in turn was based on a description of the 

mixture in a more formal document, only a portion of which was found in the files.

[23] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

In mixtures that are mostly americium by mass or where the americium atoms have been 

separated from the plutonium atoms, the americium behaves as americium, not as plutonium.

[24] Bihl, Donald E. ORAU Team. Principal Health Physicist. November 2004.

Information in the previous five sentences was based on a query of the REX database.

[25] Bihl, Donald E. ORAU Team. Principal Health Physicist. November 2004.

No information was found to indicate whether the units were activity per sample or activity per 

liter during 1964. Guidance was required; therefore, the units used in 1967 to 1969 were 

assumed.

[26] Bihl, Donald E. ORAU Team. Principal Health Physicist. November 2004.

This section is a continuation of the discussion about the 41 analyses introduced in the 

previous paragraph. This discussion is leading to an estimation of the MDA for the americium 

analysis in 1964.
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 80 of 180

 

[27] Bihl, Donald E. ORAU Team. Principal Health Physicist. November 2004.

This is based on calculations by the author on the 1964 dataset in an attempt to determine the 

MDA of the procedure at that time.

[28] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

Most of the information in this table is discussed in the text. The values during 1990 and 1991 

resulted from the use of several DOE laboratories after the default of the contract with the 

commercial laboratory. The values from October 1991 are the same as for the plutonium 

analysis. In addition, see Table 5-2, the discussion in the plutonium section, and Attribution 12.

[29] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience, results of a query of the REX database, and research 

on some of the cases in the files.

[30] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience.

[31] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC [1964], Heid [1974], DOE [1979], Simpson [1983], DOE [1985, 

1987], and Acker [1992]. 

[32] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC [1964], Heid [1974], DOE [1979], Simpson [1983], DOE [1985, 

1987], and Acker [1992]. 

[33] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC [1964], Heid [1974], DOE [1979], BMI [1982], Simpson [1983], DOE 

[1985, 1987], Acker [1992], and BMI [2003b, Appendix B]. 

[34] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was provided by [name redacted], the [title redacted], and [name redacted, 

[title redacted] in the Hanford Internal Dosimetry Program. It is also discussed in a letter to the 

Hanford Internal Dosimetry Program file [Antonio 2002]. 

[35] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC [1964], Lardy [1970], Heid [1974], DOE [1979], BMI [1982], Simpson 

[1983], DOE [1985, 1987], and Acker [1992]. 

[36] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC [1964], Lardy [1970], Heid [1974], DOE [1979], BMI [1982], Simpson 

[1983], DOE [1985, 1987], and Acker [1992]. 

[37] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience. It is also evident from a query on the dosimetry 

database. It was standard practice to include the tritium doses as external dose until the mid1980s.
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[38] Bihl, Donald E. ORAU Team. Principal Health Physicist. January 2009.

Based on a conversation with [name redacted], who worked as a [title redacted] for a time at 

the 108-B Tritium Extraction Facility.

[39] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC [1964], Lardy [1970], Heid [1974], DOE [1979], BMI [1982], Simpson 

[1983], DOE [1985, 1987], and Adkins [1992]. 

[40] Bihl, Donald E. ORAU Team. Principal Health Physicist. January 2009

Based on a conversation with [name redacted], who worked as a [title redacted] for a time at 

the 108-B Tritium Extraction Facility.

[41] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

Plutonium impurity would have resulted from incomplete separation of plutonium from the 

irradiated uranium, of which 239Pu was the isotope with the highest activity. The thorium 

impurity was probably a mix of 232Th, 228Th, and 234Th because all three isotopes were in 

PUREX; however, the assumption of 100% 232Th is efficient and favorable to claimants.

[42] Bihl, Donald E. ORAU Team. Principal Health Physicist. November 2006.

Because both the DOE Hanford RU document [DOE 2000] and the Hanford internal dosimetry 

manual [BMI 2003a] express the entity as 95Zr/95Nb, it is hard to know for sure if the activity is 

for each radionuclide or a total of both; however, DOE [2000] also refers to 106Ru/106Rh. In the 

latter case it is clear that the activity refers to the sum of the parent and progeny, so it was 

deemed reasonable to assume that the 20 nCi of 95Zr/95Nb also refers to the sum of the parent 

and progeny. Because the radionuclides are usually found close to equilibrium, the 

recommendation was made to use 10 nCi 95Zr and 10 nCi 95Nb.

[43] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

Hanford Internal Dosimetry Program staff member M. Long did some research on uranium 

discharged to the soil by asking various Hanford and PNNL soil chemists. Long concluded that 

uranium would be in a carbonate complex that most likely would be quite soluble [Long 1993]. 

[44] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience.

[45] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

For some reason the reporting of activity in the 233/234U energy region was set up as 233U early 

on; the switch to reporting as 234U was personally negotiated with the laboratory by Bihl.

[46] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

The rapid method was coded QUS for “quick uranium soluble.” Sula et al. [1989, Appendix C] 

refers to it as the “less sensitive method.” The kit code for the QUS was 7.

[47] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

The rapid method was coded QUS for “quick uranium soluble.” Sula et al. [1989, Appendix C] 

refers to it as the “less sensitive method.” The kit code for the QUS was 7.

[48] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC [1964], Lardy [1970], Heid [1974], DOE [1979], BMI [1982], Simpson 

[1983], DOE [1985, 1987], and Adkins [1992]. 
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[49] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC [1964], Lardy [1970], Heid [1974], DOE [1979], BMI [1982], Simpson 

[1983], DOE [1985, 1987], and Adkins [1992]. 

[50] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2005.

The 233U results were recorded on the same data sheets as for regular uranium analyses, and 

the units were printed on the form. However, the magnitude of the results argues against units 

of micrograms per liter, and normal uranium analysis methods would not have produced 

results that small (ranging from 2 × 10-7 to 4 × 10-6). A decision had to be made as to the units 

for the dose reconstructors to apply. It was reasoned that autoradiography was used to 

measure the tracks from the 233U alpha particles in the same manner as was being used for 

plutonium bioassay. As such, the units would have been microcuries per sample. It is likely 

someone forgot to mark the special units on the uranium bioassay form.

[51] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is evident from a query on the bioassay data that shows use of the fission product 

urinalysis from 1960 to 1964. Microfiche records of several of the workers with fission product 

urinalysis were reviewed and showed evidence of working at B Plant and Semi-Works. The 

same workers showed 90Sr urinalyses beginning in 1965. Therefore, it was concluded that the 

use of the fission product urinalysis after 1960 was for workers potentially exposed to 

radiostrontium.

[52] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

Data in the table was compiled from the references in the text and from a query of the 

bioassay database, which showed the distinct reporting levels used at different times.

[53] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This was explained to D. Bihl verbally during a regular contractual meeting between the 

Hanford Internal Dosimetry Program representative, the Battelle contract representative, and 

staff from the bioassay laboratory in about the late 1980s. This practice was considered 

standard, had been going for quite a while, and continued throughout the duration of the 

contract.

[54] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

No use of strontium titanate at Hanford has been discovered.

[55] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC [1964], Lardy [1970], Heid [1974], DOE [1979], BMI [1982], Simpson 

[1983], DOE [1985, 1987], and Adkins [1992]. 

[56] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information is based on query of the database and review of the data.

[57] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC [1964], Lardy [1970], Heid [1975], DOE [1979], BMI [1982], Simpson 

[1983], DOE [1985, 1987], and Adkins [1992]. 
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[58] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC [1964], Lardy [1970], Heid [1974], DOE [1979], BMI [1982], Simpson 

[1983], DOE [1985, 1987], and Adkins [1992]. 

[59] Bihl, Donald E. ORAU Team. Principal Health Physicist. March 2009.

Only four records of 147Pm in vivo counts were found in the REX database. It is possible that 

some early counts were not entered into the database, but it is unlikely that this type of count 

was performed other than rarely.

[60] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC [1964], Lardy [1970], Heid [1974], DOE [1979], BMI [1982], Simpson 

[1983], DOE [1985, 1987], and Adkins [1992]. 

[61] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC [1964], Lardy [1970], Heid [1974], DOE [1979], BMI [1982], Simpson 

[1983], DOE [1985, 1987], and Adkins [1992]. 

[62] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC [1964], Lardy [1970], Heid [1974], DOE [1979], BMI [1982], Simpson 

[1983], DOE [1985, 1987], and Adkins [1992]. 

[63] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2005.

This is based on queries of the REX database.

[64] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2005.

This is based on queries of the REX database.

[65] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on discussion with the [title redacted], [name redacted], a PNNL employee.

[66] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on a casual discussion with [name redacted], a retired [title redacted], now 

deceased.

[67] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

D. Bihl had access to a series of Laboratory Record Notebooks associated with the in vivo

counting facility. Laboratory Record Notebooks are proprietary and permission was not sought 

to be able to copy them.

[68] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

The earliest detection levels came from a slide in a training class on whole-body counting at 

Hanford from about 1964. The author was not stated, although it was most likely E. H. Palmer, 

the In Vivo Monitoring Program Manager at the time [Whole body counter 1964]. Some of the 

detection levels were also shown, with calculations, in Palmer’s 1960 Laboratory Record 

Notebook. Based on the calculations, D. Bihl estimated that the recorded detection levels most 

closely approximated the modern-day concept of decision level. The 10-nCi recording level for 

some of the radionuclides came from Non-Routine Whole Body Counter Calculations [Glenn 

1968]. 
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[69] Bihl, Donald E. ORAU Team. Principal Health Physicist. July 2005.

This is based on query of the REX database which shows a clear cutoff of reported results at 

0.66 nCi for 137Cs.

[70] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience. It is also evident from query on the REX database that 

shows the start of reporting of negative numbers on March 29, 1993.

[71] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

The reporting of these radionuclides is evident from a query on the database. It was also 

obvious from hard-copy printouts of in vivo data from the first computer databases that were 

used by the in vivo counting program. A yearly printout was made of all the counts for each 

calendar year and stored in the Radiological Records library. These printouts had hard-coded 

columns for 137Cs, 24Na, 65Zn, and 40K, whereas other radionuclides were noted in a comments 

field. For reasons not understood by D. Bihl, there were some counts for which one or more of 

the four regularly reported radionuclides were missing. As a general rule they were nearly 

always reported, but clearly there are some exceptions in the database.

[72] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

Although the persistence of burdens of 24Na and 65Zn were discovered when whole-body 

counting became routine and the sources were investigated over the next couple of years, it is 

favorable to claimants and reasonable to assume that the radionuclides were in the drinking 

water starting soon after the first once-through-cooling reactor started at Hanford.

[73] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal discussions with former [titles redacted], [name redacted] and 

[name redacted]. [name redacted] was the [title redacted] for many years, and [name 

redacted] was a [title redacted] in the program for many years. Both are retired. The GOK 

region is also shown on the early whole-body counting results forms.

[74] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

Because nearly every U.S. resident had detectable burdens of 137Cs in the 1960s and 1970s, 

the next three bullets were provided as guidance on when to assign occupational intakes of 

137Cs

[75] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal discussions with [name redacted] and [name redacted].

[76] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience and discussion with [name redacted]. In addition, see for 

instance the 1992 In Vivo Bioassay Statement of Work [Lynch et al. 1992, p. 21]. 

[77] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on discussion with [titles redacted] [name redacted] and [name redacted]. In 

addition, it is implied in most of the annual reports, such as Lyon et al. [1988] and others.

[78] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience and discussion with [name redacted].

[79] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2005.

This is based on queries on the REX database for 131I results.
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[80] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2005.

This is based on discussion with [name redacted] and [name redacted].

[81] Bihl, Donald E. ORAU Team. Principal Health Physicist. November 2006.

This is based on calculations by R. Traub; however, similar values can be extracted from the 

graph in West [1965]. 

[82] Bihl, Donald E. ORAU Team. Principal Health Physicist. June 2005.

The terms “detection limit” or “less-than” during the 1960s appear to be most closely 

associated with the present concept of decision level. The MDA is approximately twice the 

decision level; therefore, the recommendation was made to assume that the MDA was twice 

the reported detection level.

[83] Bihl, Donald E. ORAU Team. Principal Health Physicist. November 2006.

This is based on discussion with [name redacted]; however, this is an educated guess by 

[name redacted] based on present practices and his knowledge of the detection systems at 

the time. Documentation that the 232Th body burdens in the 1960s and 1970s were based on 

228Ac has not been found. It is also not known if an adjustment for disequilibrium between 228Ac 

and 232Th was made before the results were reported. It was considered favorable to claimants 

to assume the disequilibrium adjustment was not made and to direct dose reconstructors to 

revise the 232Th results based on disequilibrium factors in the text.

[84] Traub, Richard. ORAU Team. Principal Health Physicist. November 2006.

This is based on calculations by R. Traub using MATLAB. The calculations incorporated the 

latest ICRP biokinetic models with independent systemic biokinetics for the major progeny.

[85] Traub, Richard. ORAU Team. Principal Health Physicist. November 2006.

This is based on calculations by R. Traub using MATLAB. The calculations incorporated the 

latest ICRP biokinetic models with independent systemic biokinetics for the major progeny.

[86] Bihl, Donald E. ORAU Team. Principal Health Physicist. December 2006.

Because 228Ac is usually considered part of natural background, it is not routinely calculated or 

reported in a whole-body count spectrum. If it had been known that the worker was exposed to 

thorium, any detectable activity would have been calculated and reported, but otherwise it 

would take a more significant peak to prompt an investigation about possible thorium 

exposure. The exact level of activity that would have caused an investigation has not been 

found; the 2-nCi level was an assumption based on the fact that levels well below 1 nCi were 

being recorded in the early 1970s and detection levels have generally improved over time.

Therefore, 2 nCi is favorable to claimants.

[87] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience and also evident from a query of the REX database.

[88] Traub, Richard. ORAU Team. Principal Health Physicist. December 2006.

These were calculated by R. Traub using MATLAB. The calculations incorporated the latest 

ICRP biokinetic models with independent systemic biokinetics for the major progeny.

[89] Traub, Richard. ORAU Team. Principal Health Physicist. December 2006.

These were calculated by R. Traub using MATLAB. The calculations incorporated the latest 

ICRP biokinetic models with independent systemic biokinetics for the major progeny.
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[90] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience and discussion with [name redacted], [title redacted]. 

The correction for shine from the liver is also mentioned in BMI [1990]. 

[91] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on review of the REX database and a sampling of whole-body count results 

forms in the 1960s and 1970s.

[92] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on discussion with [name redacted], [title redacted]. Document research has not 

found a reference that states this precisely, but Lynch [1992] shows the detailed work being 

done in relation to decision levels at that time.

[93] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is evident from a review of various whole-body count results forms.

[94] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

These changes were evident from review of the REX database.

[95] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

Implementation of NEXEC is discussed in Lyon et al. [1996]. BMI [1995, Table 4.1] shows the

12 regions for the stand-up counter, and Table 4.5 shows the 20 regions for the coaxial 

germanium counter.

[96] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on personal experience and discussion with [name redacted].

[97] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

The minutes of the November 1998 meeting of the Dosimetry Advisory Committee [Bihl 1998] 

show the proposal by T. Lynch to reduce the routinely reported radionuclides in whole-body 

counts to 137C, 60Co, 154Eu, 40K, and 24Na. By the time the ABACOS software was implemented 

in November 1999, 24Na had been dropped because Hanford contractors believed there was 

no significant source of 24Na left. Subsequent minutes of the Dosimetry Advisory Committee 

were scanned, but documentation of the decision to drop 24Na was not found.

[98] Bihl, Donald E. ORAU Team. Principal Health Physicist. May 2006.

The FFTF was sodium cooled, so neutron activation of the sodium was a potential source of 

contamination.

[99] Bihl, Donald E. ORAU Team. Principal Health Physicist. March 2007.

The principal 154Eu photopeak is at 1.27 MeV with 35% abundance; the photopeak that was 

used to quantify 60Co was at 1.33 MeV with 100% abundance. The detector efficiencies per 

emitted photon would have been nearly the same for the two photons; the background counts 

in the respective energy regions would also have been similar. Therefore, the MDAs would 

have differed by the ratio of the yields, or 1.00/0.35 ≈ 3.

[100] Napier, Bruce A. ORAU Team. Principal Health Physicist. December 2008.

B. Napier conducted a study of present alpha to beta-gamma emitting radionuclides in all the 

high-level waste tanks from the TWINS 2001 Tank Farm Inventory Database. The ratios in the 

tanks were fit to a lognormal distribution and the 95th-percentile ratio was obtained. The ratios 

from the TWINS database were compared to a database for the Hanford Environmental Dose 

Reconstruction Project [Napier 1992] after correction for removal of strontium and cesium at B 

Plant and the West Encapsulation Facility and after decay correction to 2001. The study 
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showed that in general the tanks historically had smaller alpha to beta/gamma ratios because 

the fuel being processed had less cooling time and, of course, present waste has had many 

additional years of decay. Shorter-cooled and shorter-decayed fuel has more activity from 

beta- and gamma-emitting radionuclides.

[101] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is evident from a query on the database. Error limits are listed as “±10% at the 95% 

confidence level at 10 times the detection limits” in the earliest statement of work for the 

bioassay laboratory [AEC 1964], and similar statements show up in subsequent statements of 

work, but uncertainties for each sample were not reported at that time. There was probably a 

change to the statement of work in 1981 that instigated or accompanied the change in that 

year to start reporting uncertainties for each sample, but a copy of that statement of work has 

not been found. Propagation and reporting of uncertainty for each sample shows up in the 

1983 bioassay laboratory statement of work [Simpson 1983] and in each statement of work 

thereafter.

[102] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

The collection instructions are to collect starting with the last excretion before retiring for sleep, 

any other voidings during the night (or sleep period if on shift), and first excretion on rising in 

the morning for two consecutive nights. This covers about 18 hours or about 75% of a 24-hour 

period.

[103] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on discussion with the [title redacted] [name redacted].

[104] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on discussion with the [title redacted] [name redacted].

[105] Bihl, Donald. ORAU Team. Principal Health Physicist. October 2003.

This is based on discussion with the [title redacted] [name redacted].

[106] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

The 95th-percentile phantom did not exist during the period that the 1-m arc or the shadow 

shield detectors were in use, and that phantom was not measured on the standup counter, so 

the impact of large workers on results from those detectors has not been quantified. T. P. 

Lynch believed the 25% bias was a reasonable figure to use for the latter detectors as well as 

the coaxial counter based on his general experience with the various detectors.

[107] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

This is based on discussion with the [titles redacted] [name redacted] and [name redacted].

[108] Bihl, Donald E. ORAU Team. Principal Health Physicist. December 2005.

N Reactor, the last source of short-lived fission products, shut down in January 1987 after a 

brief operating period, and was also shut down for most of 1986. The December 31, 1988, 

date was arbitrarily chosen to allow time for the short half-life radionuclides to decay until they 

were no longer significant contributors to contamination.

[109] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

Workers without at least a basic dosimeter were not allowed into radiation areas or general 

areas of the site (200 Area, 300 Area, etc.) where there was some risk of exposure. Some 

companies gave beginning work or termination bioassay to everyone regardless of their work 

location or tasks, and those bioassays are not indicative of exposure. Therefore, after 

consultation with other ORAU Team experts (including J. Fix, E. Scalsky, and E. Brackett), the 
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conclusion was reached that workers without at least a basic dosimeter were not at risk for 

occupational intakes, and the instruction was made to assign environmental intakes only.

[110] Bihl, Donald E. ORAU Team. Principal Health Physicist. December 2005.

This is based on queries of the REX database and an incident record in the worker’s file about 

the T Plant incident.

[111] Bihl, Donald E. ORAU Team. Principal Health Physicist. December 2005.

Equilibrium thyroid burden from daily inhalation of 20,000 pCi of type F 131I is 29.5 nCi after 

about 80 days. This calculation was performed using IMBA. This burden exceeds 93% of all 

the measurements made on at-risk workers in 1945 and 1946 and exceeds 90% of the 

measurements made from 1962 to 1969, so it was deemed to be favorable to claimants to 

apply the dose to unmonitored workers in years before implementation of routine in vivo 

counts.

[112] Bihl, Donald E. ORAU Team. Principal Health Physicist. December 2005.

The whole-body burden from daily inhalation of 3,400 pCi of type F 131I is 5 nCi after only 

10 days of intake and reaches an equilibrium burden of 7.2 nCi after about 80 days. This 

calculation was performed using IMBA.

[113] Bihl, Donald E. ORAU Team. Principal Health Physicist. December 2005.

N Reactor shut down in May 1986 and restarted very briefly in January 1987. With an 8-day 

half-life, 131I activities would have been negligible by the end of 1987.

[114] Bihl, Donald E. ORAU Team. Principal Health Physicist. May 2006.

This is based on query of the REX database.

[115] Bihl, Donald E. ORAU Team. Principal Health Physicist. May 2006.

This is based on query of the REX database.

[116] Bihl, Donald E. ORAU Team. Principal Health Physicist. Joseph Lochamy. XCEL Engineering.

Scientist. Richard Sparks. CDE Dosimetry Services. Principal Health Physicist. December 

2006.

The tritium dose data is from a query on REX by D. Bihl, a lognormal statistical analysis was 

performed by Lochamy in accordance with ORAUT-PROC-0095, Generating Summary 

Statistics for Coworker Bioassay Data [ORAUT 2006], and a validation of statistical analysis 

was performed by R. Sparks. 

[117] Bihl, Donald E. ORAU Team. Principal Health Physicist. May 2006.

This was calculated using IMBA.

[118] Bihl, Donald E. ORAU Team. Principal Health Physicist. May 2006.

There were considerable differences in potential exposure to tritium between the 108-B facility, 

which involved extracting and packaging curie quantities of tritium in a relatively small space, 

and exposures at large, heavily ventilated facilities with unconcentrated tritium, such as the 

reactors and fuel separation facilities. Most of the tritium at the latter facilities was exhausted 

through the stacks or discharged to the ground in cribs. It is likely that the relative exposure to 

workers at the reactors or fuel separations plants was less than an order of magnitude relative 

to exposure at 108-B. However, lacking air sample data, the order of magnitude ratio was 

considered reasonable and favorable to claimants.

[119] Bihl, Donald E. ORAU Team. Principal Health Physicist. May 2006.

There were considerable differences in potential exposure to tritium between the 108-B facility, 
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which involved extracting and packaging curie quantities of tritium in a relatively small space, 

and exposures at large, heavily ventilated facilities with unconcentrated tritium, such as the 

reactors and fuel separation facilities. Most of the tritium at the latter facilities was exhausted 

through the stacks or discharged to the ground in cribs. It is likely that the relative exposure to 

workers at the reactors or fuel separations plants was less than an order of magnitude relative 

to exposure at 108-B. However, lacking air sample data, the order of magnitude ratio was 

considered reasonable and favorable to claimants.

[120] Bihl, Donald E. ORAU Team. Principal Health Physicist. November 2006.

This was calculated using IMBA.

[121] Bihl, Donald E. ORAU Team. Principal Health Physicist. Joseph Lochamy. Scientist. XCEL 

Engineering, Inc. Richard Sparks. CDE Dosimetry Services, Inc. Principal Health Physicist.

December 2006.

The tritium dose data is from a query on REX by D. Bihl, a lognormal statistical analysis was 

performed by J. Lochamy in accordance with ORAUT-PROC-0095, Generating Summary 

Statistics for Coworker Bioassay Data [ORAUT 2006], and a validation of statistical analysis 

was performed by R. Sparks.

[122] Bihl, Donald E. ORAU Team. Principal Health Physicist. November 2006.

This was calculated using IMBA.

[123] Bihl, Donald E. ORAU Team. Principal Health Physicist. November 2006.

Although the tritium source term would have been smaller in the late 1980s and early 1990 

because of reduced processing of fuel, it is favorable to claimants to use the values for 1984 to 

1986 for the later years.

[124] Bihl, Donald E. ORAU Team. Principal Health Physicist. June 2005.

These were calculated using IMBA. Ingestion of 40,000 pCi/d produces approximately 

1 mrem/yr to all organs.

[125] Bihl, Donald E. ORAU Team. Principal Health Physicist. June 2006.

This is a summary of information developed in the text.

[126] Bihl, Donald E. ORAU Team. Principal Health Physicist. June 2006.

This is based on discussion with the [title redacted] [name redacted]. It reflects radionuclides 

being encountered during D&D operations of old facilities, burial sites, liquid effluent discharge 

zones, etc. Because most of the work is outside, the contractor uses lapel air samplers to 

monitor for intakes. The 40-DAC-hr value comes from the requirement to initiate special 

bioassay if an individual’s cumulative intake in a year as measured by the air samplers 

reaches or exceeds 40-DAC-hr [BMI 2003b]. 

[127] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

The air concentration tolerance value as applied to fission products was deduced from the fact 

that other values were given for “product,” 131I, and uranium. In addition, the magnitude of the 

limit implies general fission products.

[128] Bihl, Donald E. ORAU Team. Principal Health Physicist. October 2003.

The air concentration tolerance value as applied to fission products was deduced from the fact 

that other values were given for “product,” 131I, and uranium. In addition, the magnitude of the 

limit implies general fission products.
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 90 of 180

[129] Bihl, Donald E. ORAU Team. Principal Health Physicist.

This is from a query on the REX database by the REX custodian.

[130] Arno, Matthew. ORAU Team. Dose Reconstructor. June 11, 2007.

The periods for how data was recorded are taken directly from examination of the Health 

Related Energy Research Branch (HERB) database.

[131] Bihl, Donald E. ORAU Team. Principal Health Physicist. July 2005.

The uranium intakes were identified by their pattern of excretion associated with each 

individual followed by review of each individual’s radiological exposure file, to which D. Bihl 

had access. The description in the radiological exposure file was used to determine the 

specifics of the intakes. Most intakes investigated in this way were not excluded from the 

database; only intakes considered unusual and not relevant to a co-exposure analysis were 

excluded.

[132] Bihl, Donald E. ORAU Team. Principal Health Physicist. July 2005.

The promethium intakes were identified by the patterns of excretion that were associated with 

each individual followed by review of each individual’s radiological exposure file to which D. 

Bihl had access. The description in the radiological exposure file was used to determine the 

specifics of the intakes. Most intakes investigated in this way were not excluded from the 

database; only intakes considered unusual and not relevant to a co-exposure analysis were 

excluded.

[133] Arno, Matthew. ORAU Team. Dose Reconstructor. June 2007.

Samples with known volumes are normalized using a urinary excretion rate of 1.4 L/d.

Samples without a known volume are assumed to be representative of a full day of urinary 

excretion because no additional information is available.

[134] Arno, Matthew. ORAU Team. Dose Reconstructor. June 2007.

Ten-year aged fuel-grade plutonium is the default assumption according to the main text of 

this document. However, before 1955, it was not possible to have 10-year aged plutonium.

Therefore, the most favorable to claimant credible mixture is used.

[135] Arno, Matthew. ORAU Team. Dose Reconstructor. June 2007.

Internal dose conversion factors for 234U, 235U, and 238U are all similar, but those for 234U are 

the most favorable to claimants and are therefore used as a simplification.

[136] Bihl, Donald E. ORAU Team. Principal Health Physicist. July 2005.

Plot created by D. Bihl from data in Heeb [1994, Appendix A].

[137] Bihl, Donald E. ORAU Team. Principal Health Physicist. March 2009.

The Hanford Site Description [ORAUT 2010a] lists the buildings and periods of work with 

mixtures highly enriched in 147Pm or with purified 147Pm. In general, the early work was in 

developing separations techniques and was done by only a few chemists and chemical 

technicians, and major production started in 1966. Most of the work was done in the hot cells 

in the 325 Building with early low-level work in fume hoods in the 325 Building or the 

1F cubicle in 222-S. Urinalyses were taken for 29 workers in the 1973 contamination spread in 

the 325 Building [AEC 1963], but only the calculated body burdens were found. These were 

converted to intakes and urine excretion using the Langham equation as stated in the report.

The calculated urine excretion values were then recalculated to intakes using IMBA and 

present-day models. The 95th percentile of these recalculated intakes were comparable, albeit 

slightly higher, compared to annualized intakes from Table C-32 for the period of major 

production from 1966 to 1969. Considering that the February 1963 bioassays resulted from a 
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 91 of 180

 

major contamination spread, and that work conditions were similar in the two periods, it was 

concluded that the 1966 to 1969 co-exposure intakes were reasonable surrogates for intakes 

by workers in the 1961 to 1965 period that is not associated with contaminating events.
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 92 of 180

 

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GLOSSARY

absorbed dose

Amount of energy (ergs or joules) deposited in a substance by ionizing radiation per unit mass 

(grams or kilograms) of the substance and measured in units of rads or grays. See dose.

active

Term at Hanford in the 1940s and 1950s for radioactive. Example, “production of acetylene 

from the active water, with subsequent measurement of the ionization caused by the tritium 

beta particle.” See radioactive.

age or aging

In relation to reactor fuel and mixtures of plutonium isotopes, time since the step in the 

refinement process that separates americium from the mixture.

alpha particle

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.

ashing

Driving off of all moisture in a sample. Accomplished by heating the sample in an oven or by 

dissolving it in a liquid (often nitric acid), then evaporating to dryness. The latter technique is 

called wet ashing.

bremsstrahlung

Electromagnetic radiation released as a result of inelastic scattering of a moving charged 

particle within the nucleus of an atom. X-rays produced in a typical medical X-ray tube 

frequently originate from inelastic scattering of accelerated electrons in the anode material.

cooling

In relation to reactor fuel, time since removal of the fuel from the reactor core.

curie (Ci)

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.

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 rad, rep, or grays.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See film dosimeter.

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 
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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. 

Energy Employees Occupational Illness Compensation Program Act of 2000 (EEOICPA; 

42 USC 7384-7385) 

Law that provides for evaluation of cause and potential compensation for energy employees 

who have certain types of cancer.

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.

film dosimeter

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

fission

Splitting of the nucleus of an atom (usually of a heavy element) into at least two other nuclei 

and the release of a relatively large amount of energy. This transformation usually releases 

two or three neutrons. 

fuel-grade plutonium

Plutonium isotopic mixture with approximately 12% 240Pu. 

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, particle, or photon (γ)

See gamma radiation. 

glovebox

Enclosure with special rubber gloves through which an operator can handle radioactive or 

toxic material without risk of injury or contamination, normally operated at a slightly reduced 

pressure so that air leakage, if any, is inward.

in vitro bioassay

Measurements to determine the presence of or to estimate the amount of radioactive material 

in the excreta or in other biological materials removed from the body.

in vivo bioassay

Measurements of radioactive material in the human body using instrumentation that detects 

radiation emitted from the radioactive material in the body.
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ionization

Process by which a neutral atom or molecule loses or gains electrons and therefore acquires a 

net electrical charge.

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.

National Institute for Occupational Safety and Health (NIOSH)

U.S. agency responsible for dose reconstruction under the Energy Employees Occupational 

Illness Compensation Program Act of 2000. Part of the Centers for Disease Control and 

Prevention, which is part of the U.S. Department of Health and Human Services, NIOSH is the 

federal agency responsible for conducting research and making recommendations for the 

prevention of work-related injury and illness.

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.

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.

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.

probability of causation

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Program Act of 2000, the percent likelihood, at the 99th percentile, that a 

worker incurred a particular cancer from occupational exposure to 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. See radionuclide.

radionuclide

Radioactive nuclide. See nuclide.
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reactor

Device in which a fission chain reaction occurs under controlled conditions to produce heat or 

useful radiation for experimental purposes or to generate electrical power or nuclear fuel.

reliably detectable activity (RDA)

Measure for in vivo counting in the 1980s defined as three standard deviations of the spectral 

continuum with a peak discernable by the naked eye.

rep

Historical quantity of radiation (usually other than X-ray or gamma radiation) originally defined 

as 93 ergs absorbed per gram in the body and redefined in the 1940s or early 1950s as the 

amount that would liberate the same amount of energy (93 ergs per gram) as 1 roentgen of 

X- or gamma rays. Replaced by the gray in the International System of Units; 1 rep is 

approximately equal to 9.3 milligray. The word derives from roentgen equivalent physical; rep 

is also the plural.

Special Exposure Cohort (SEC) [42 USC 7384]

[A] “member of the Special Exposure Cohort” means a Department of Energy employee, 

Department of Energy contractor employee, or atomic weapons employee who meets any of 

the following requirements:

(A) The employee was so employed for a number of work days aggregating at least 250 

work days before February 1, 1992, at a gaseous diffusion plant located in Paducah, 

Kentucky, Portsmouth, Ohio, or Oak Ridge, Tennessee, and, during such 

employment— 

(i) was monitored through the use of dosimetry badges for exposure at the plant of 

the external parts of employee’s body to radiation; or

(ii) worked in a job that had exposures comparable to a job that is or was monitored 

through the use of dosimetry badges.

(B) The employee was so employed before January 1, 1974, by the Department of 

Energy or a Department of Energy contractor or subcontractor on Amchitka Island, 

Alaska, and was exposed to ionizing radiation in the performance of duty related to 

the Long Shot, Milrow, or Cannikin underground nuclear tests.

(C) (i) Subject to clause (ii), the employee is an individual designated as a member of 

the Special Exposure Cohort by the President for purposes of the compensation 

program under section 7384q of this title.

(ii) A designation under clause (i) shall, unless Congress otherwise provides, take 

effect on the date that is 180 days after the date on which the President submits to 

Congress a report identifying the individuals covered by the designation and 

describing the criteria used in designating those individuals.

thoron

Informal name for 220Rn.

U.S. Atomic Energy Commission (AEC)

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The U.S. Energy Research and Development 
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Administration and the U.S. Nuclear Regulatory Commission assumed separate duties from 

the AEC in 1974. The U.S. Department of Energy succeeded the U.S. Energy Research and 

Development Administration in 1979.

weapons-grade plutonium

Plutonium isotopic mixture with nominally 4% to 6% 240Pu. 

whole-body counter

Equipment used to perform in vivo bioassay. Radiation emitted from radioactive material 

deposited throughout the body is measured.

X-ray

See X-ray radiation. 

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 

TOLERANCE DOSE AND TOLERANCE VALUES

Various health physics textbooks and histories document the history of the tolerance dose. The 

description below was extracted from Radiation Protection: A Guide for Scientists, Regulators, and 

Physicians, Fourth Edition [Shapiro 2002] and from Selected Materials on Radiation Protection 

Criteria and Standards: Their Basis and Use [Parker 1960]. H. M. Parker was the manager of the 

radiation protection activities at the start of Hanford and eventually became the manager of the 

Hanford Laboratories.

The tolerance dose was based on a rate of radiation exposure that produced no obvious harm. The 

basis for harm changed as understanding of the health effects of radiation improved.

The first tolerance dose was established in 1934 by the International X-Ray and Radium Protection 

Commission as 0.2 R/d. In 1936 the U.S. Advisory Committee on X-Ray and Radium Protection 

reduced the tolerance dose to 0.1 R/d. This value was used at the Manhattan Project sites during 

World War II. For instance, Parker [1960, p. 6] stated that in 1945 the values accepted as a working 

basis for occupational exposure were:

• 100 mR/d for external X- and gamma radiation

• 10 to 14 Ci/cm3 for radon in air of working rooms, and

• 0.1 μg of radium as the maximum allowable amount deposited in the body.

In 1949, the NCRP recommended reduction of the permissible dose to 0.3 rem/wk. In 1953, the 

National Committee on Radiation Protection (predecessor of NCRP) developed maximum permissible 

amounts of certain radionuclides in the body and MPCs in air and water, which were published as 

National Bureau of Standards Handbook 52 [NBS 1953]. At that point tolerance values were replaced 

with MPBBs and MPCs.

Hanford’s radiation protection standards were in accordance with these national and international 

standards. For instance, Table A-1 lists examples of the “tolerance limits for prolonged exposure” at 

Hanford in 1945 [Cantril 1945]. 

Table A-1. Example tolerance limits, 1945.

Measured quantity Tolerance limit

External gamma and X- radiation 0.1 R/d

External beta radiation 0.1 repa/d

Fast neutron radiation 0.02 rep/d

Slow neutron radiation 0.025 rep/d

Internal alpha radiation 0.01 rep/d

Radium deposition in body 0.1 μg total accumulation

Radio concentration in atmosphereb 1 × 10-14 Ci/cm3

Radioactive I-131 in atmosphere 1.0 × 10-13 Ci/cm3

Mixed fission products in drinking water 1.2 × 10-9 Ci/cm3

Productc concentration in drinking water 10-5 μg/cm3

Product concentration in atmosphere 5 × 10-10 μg/cm3

Product deposition in body 0.5-μg total accumulation

Uranium dust in atmosphere 1.5 × 10-4 μg/cm3

a. The rep (from roentgen equivalent physical) was a unit of dose from 

particulate radiation invented by H. Parker. It was a precursor to the rad. It 

was defined as an absorbed dose of 83 ergs/g of tissue and was later 

changed to 93 erg/g.

b. Assumed to refer to fission products [127].

c. “Product” was a euphemism for plutonium.
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ATTACHMENT A

TOLERANCE DOSE AND TOLERANCE VALUES (continued)

By March 1949, the Hanford limits had changed to incorporate the 0.3-rem/wk guidance from the 

National Committee. Examples are quoted below from Patterson [1949], but these same air 

concentrations limits were discussed as Hanford limits in a lecture series by Parker [1947]. The 

Patterson letter refers to these values as permissible tolerances.

External Radiation Limits

Whole body: 0.3 rem/wk or whatever the National Committee recommended.

Hands only: 1.0 rem/wk or whatever the National Committee recommended.

Internal Emitter Limits

0.3 rem/wk to the significant organ or whatever the National Committee recommended.

Drinking Water

Uranium: 100 μg/L (believe this is high).

Plutonium: 0.01 μg/L (intend to revise to not more than 0.001 μg/L).

Mixed fission products: 0.1 μCi/L (to be changed when new figure provided by K. Z. Morgan’s 

subcommittee). (Patterson [1949] used “μc,” which has been assumed to mean microcuries.)

Air Contamination Limits

Respiratory protection was required when the following limits were exceeded or expected to be 

exceeded:

• Uranium: >0.05 μg/L

• Plutonium: >2 × 10-8 μg/L (to be revised perhaps to 2 × 10-9 μg/L)

• Mixed fission products: >10-6 μCi/L [128]

Other air contamination limits included:

• Tritium: Gas 0.1 μCi/L, vapor 0.01 μCi/L

• Carbon: 0.02 μCi/L

• Argon: 1.6 × 10-3 μCi/L

• Iodine: 1.5 × 10-6 μCi/L

• Xenon: 0.01 μCi/L

The tolerance values for air and drinking water were based on the dose to the significant organ and 

state-of-the-art understanding of the biokinetics of the element in the body. An assumption of 

continuous inhalation or drinking of water was usually made, although in the example below a 

tolerance value based on a single inhalation was also established. A sample calculation was found 

[Parker 1945] that provides the calculation of the tolerance value for “potentially long-continued 

exposure” to 131I and a “one-shot tolerance value,” the latter assuming one 8-hour exposure. For the 

chronic exposure, Parker assumed a thyroid radiation tolerance of 1 R/d, which resulted from an 

equilibrium thyroid burden of 1.95 μCi, which would result from a daily intake of 0.85 μCi. Parker 

assumed the volume of air that was breathed per 8-hour workday was 8 × 106 cm3. For the acute 

intake, Parker referenced a “Project Handbook” that condoned a dose of 100 R. However, he stated 

that this was too close to the dose that was given patients in the treatment of hyperthyroidism, so he 

arbitrarily reduced the dose by a factor of 10. He then calculated an air concentration for one 8-hour 

exposure to be 1.2 × 10-11 Ci/cm3.
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ATTACHMENT B 

CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS

LIST OF TABLES

TABLE TITLE PAGE

B-1 Codes and radionuclides associated with bioassay at Hanford..............................................111

B-2 Sample type codes .................................................................................................................114

B-3 Bioassay reason codes...........................................................................................................114

B-4 Excreta sample kit codes ........................................................................................................115

B-5 In vivo body location codes.....................................................................................................116

B-6 Excreta unit codes ..................................................................................................................116

B-7 Excreta processing codes.......................................................................................................116

B-8 Excreta laboratory codes ........................................................................................................117

B-9 Excreta no-sample codes .......................................................................................................117

B-10 In vivo invalid result codes......................................................................................................117

B-11 INTERTRAC mode-of-intake codes........................................................................................117

B-12 INTERTRAC evaluation reason codes ...................................................................................118

B-13 INTERTRAC source-of-intake codes ......................................................................................118

B-14 INTERTRAC miscellaneous codes .........................................................................................118

B-15 Special whole-body count resolution codes............................................................................118
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ATTACHMENT B 

CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS (continued)

This attachment is adapted from BMI [2003b]. These codes apply to information in the REX database 

and include present and historical uses. Different codes were implemented at different times 

according to the needs at the time; the early dates have few if any codes.

Table B-1. Codes and radionuclides associated with bioassay at Hanford [129]. 

Code Description Comment

AAAA1 Americium Probably Am-241

AAAA2 Americium Probably Am-241

AAAA3 Americium Probably Am-241

AAAA4 Americium Probably Am-241

AAAA5 Americium Probably Am-241

AAAA6 Americium Probably Am-241

AAAA7 Americium Probably Am-241

AC225 Actinium-225 None

ACS Actinium-227, thorium-227 Scheduling code

AC227 Actinium-227 None

AC228 Actinium-228 None

AG110 Silver-110 None

AM241 Americium-241 None

AM242 Americium-242 None

AM243 Americium-243 None

BA140 Barium-140 None

BETA Beta None

BI213 Bismuth-213 None

BI214 Bismuth-214 None

BK249 Berkelium-249 None

BR 82 Bromine-82 None

C 14 Carbon-14 None

CE141 Cerium-141 None

CE143 Cerium-143 None

CE144 Cerium-144 None

CF249 Californium-249 None

CM242 Curium-242 None

CM244 Curium-244 None

CO 58 Cobalt-58 None

CO 60 Cobalt-60 None

CR 51 Chromium-51 None

CS134 Cesium-134 None

CS137 Cesium-137 None

EU152 Europium-152 None

EU154 Europium-154 None

EU155 Europium-155 None

EU156 Europium-156 None

EV155 Unknown Probably a typographical error for Eu-155 that got left in the 

database 

EV156 Unknown Probably a typographical error for Eu-156 that got left in the 

database 

FE 59 Iron-59 None

FP Fission products None

GA Gross alpha None

GB Gross beta None
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 112 of 180

ATTACHMENT B

CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS (continued)

Code Description Comment

GELI Gamma-GeLi detector Excreta scheduling code for a gamma scan with a germanium 

detector

GOK Unknown See text

GS Gamma NaI detector Excreta scheduling code for a gamma scan with a NaI detector

H 3 Tritium None

I 125 Iodine-125 None

I 129 Iodine-129 None

I 131 Iodine-131 None

I 133 Iodine-133 None

IAM Isotopic americium Excreta scheduling code for americium separation and alpha 

spectrometry

ICA Unknown Probably scheduling code for americium and curium via alpha 

spectrometry

ICM Cm isotopic Excreta scheduling code for curium isotopes via alpha spectrometry

IEU Eu isotopic Excreta scheduling code for europium separation and isotopic 

analysis

IPA Isotopic Pu and Am-241 Excreta scheduling code

IPIU Isotopic Pu, isotopic U Excreta scheduling code 

IPS Isotopic Pu and Sr Excreta scheduling code

IPSA Isotopic Pu, Sr tot & Am-241 Excreta scheduling code; Sr tot means radiostrontium by gross beta

IPSR Seq Pu isotopic Sr-total Excreta scheduling code for isotopes of Pu and radiostrontium

IPU Isotopic plutonium Excreta scheduling code

IPUB Plutonium isotopic, Pu-241 Excreta scheduling code; Pu-241 separately analyzed by beta 

counting

IPUBA Pu isotopic, Pu-241, Am-241 Excreta scheduling code

IPUL Low-level isotopic Pu Pu-238 and Pu-239 using a 10,000-min count

IRA Radium isotopic Excreta scheduling code

IR192 Iridium-192 Excreta scheduling code

ISCP Sequential Sr-90, Ce, Pm Excreta scheduling code

ISPEC Gamma spectroscopy Excreta scheduling code

ISR Sr isotopic Excreta scheduling code

ITH Thorium isotopic Excreta scheduling code

ITPAC Seq isotopic Pu, Cm- & Am241Excreta scheduling code

IU U isotopic Excreta scheduling code

IUPU Isotopic plutonium/NU Excreta scheduling code

K 40 Potassium None

LA140 Lanthanum-140 None

LEPD Low-energy photon detector Excreta scheduling code for low-energy photon scan

MFP Mixed fission products None

MN 54 Manganese-54 None

MO 99 Molybdenum-99 None

NA 22 Sodium-22 None

NA 24 Sodium-24 None

NAI Gamma NaI detector Excreta scheduling code

NB 95 Niobium-95 None

NP237 Neptunium-237 None

NP239 Neptunium-239 None

PB210 Lead-210 None

PB212 Lead-212 None

PM147 Promethium-147 None

PO210 Polonium-210 None
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ATTACHMENT B

CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS (continued)

Code Description Comment

PR144 Praseodymium-144 None

PU Plutonium alpha Total alpha from Pu isotopes after separation

PUMIX Plutonium alpha Total alpha from Pu isotopes and Am-241

PU238 Plutonium-238 None

PU239 Plutonium-239 When pertaining to excreta samples, it’s actually Pu-239+240

PU240 Plutonium-240 None

PU241 Plutonium-241 None

PU242 Plutonium-242 None

QUS U Quick Uranium Soluble; excreta scheduling code for elemental U

QUS 1 U Same as QUS

QUS 2 U Same as QUS

RA224 Radium-224 None

RA225 Radium-225 None

RA226 Radium-226 None

RA228 Radium-228 None

RH106 Rhodium-106 None

RND Radon daughters None

RU103 Ruthenium-103 None

RU106 Ruthenium-106 None

S 35 Sulfur-35 None

SB124 Antimony-124 None

SB125 Antimony-125 None

SCP Sequential Sr-total Ce Pm Excreta scheduling code

SM153 Samarium-153 None

SR Strontium Total radiostrontium by beta counting

SR 89 Strontium-89 None

SR 90 Strontium-90 When pertaining to excreta samples, Sr-90 by yttrium ingrowth

TAC Total actinides None

TC 99 Technetium-99 None

TH227 Thorium-227 None

TH228 Thorium-228 None

TH230 Thorium-230 None

TH232 Thorium-232 None

TH234 Thorium-234 None

TL208 Thallium-208 None

U Elemental uranium None

URAN Elemental uranium None

U DEP DU None

U NAT NU None

U 233 Uranium-233 See uranium discussion in text

U 234 Uranium-234 Actually U-234+233, but usually U-234 

U 235 Uranium-235 None

U 236 Uranium-236 None

U 238 Uranium-238 None

UMIX Uranium mix Total uranium, used for intakes not bioassay

UMS U-235 U-236 U-238 U-234 None

US U None

XX 0 Isotope will have no result None

ZN 65 Zinc-65 None

ZR 95 Zirconium-95 None
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ATTACHMENT B 

CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS (continued)

Table B-2. Sample type codes. 

Code Type of sample

B Blood

F Feces

S Sputum

T Tissue

U Urine

Table B-3. Bioassay reason codes. 

Code Name Description

BL Baseline Measurement performed to establish reference level against which 

subsequent measurements will be compared. In general, could be for 

new employees or for established employees before commencing work 

with radioactive materials, beginning specific type of radiation zone work, 

or making offsite trip where potential intakes could occur.

PR Periodic Measurement performed at regularly scheduled interval.

EA End of Assignment Measurement performed after completion of specific work assignment, 

but not end of employment.

SP Special Measurement performed as part of specific investigation of potential 

internal dose. Could include response to off-normal work conditions or 

followup of abnormal periodic measurements.

CR Contractor Request Measurement requested by employer for reasons other than periodic, 

baseline, end of assignment, or special investigation.

RA Reanalysis A First reanalysis of sample by taking another aliquot and repeating same 

radiochemical or chemical analysis.

RB Reanalysis B Second reanalysis of sample by taking another aliquot and repeating 

same radiochemical or chemical analysis.

R1 Recount 1 First recount of original excreta sample or repeat in vivo examination.

R2 Recount 2 Second recount of original excreta sample or repeat in vivo examination.

QR Quality and Research Measurement performed as part of quality control, quality assurance, or 

research work.

TM or TS Termination Final bioassay at termination of employment.

12 Contract Work In vivo measurement performed under contract to customers rather than 

for Hanford employees.

20 Source Count In vivo source count made for system calibration or as function check, 

usually using known check source.

30 Background Count In vivo system background measurement performed for system 

calibration or as functional check.
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ATTACHMENT B 

CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS (continued)

Table B-4. Excreta sample kit codes.

a

Kit codeb

D/R, P/U Media Sample description

1, P Urine Approximate 24-hr urine collection. Collected at home over 2-d period.

Used for routine sampling and when larger volume sample is desired.

Designated sample date is day after kit delivery to employee.

2, Q Urine Approximate 12-hr urine collection for termination sampling only. Collected 

at home overnight. Designated sample date is day after date of kit delivery 

to employee.

3, R Urine Total 24-hr urine collection. Collected at home and at work (if necessary) 

to collect all urine voided during 24-hr period. Generally used for sampling 

immediately following occurrence or for work restriction sampling.

Designated sample date is day after delivery or date on which sample 

collection began.

4, S Urine Single void (spot urine) collection. Collection in single bottle, used for initial 

indications of intake. Designated sample date is date of voiding.

5, T Feces Collection of single fecal voiding usually for investigation of potential 

intake. Sample date is day after kit delivery or date on which sample was 

actually voided.

6, U Urine Partial day or approximate 12-hr collection. Usually collected at home 

overnight. Used for collection following occurrence or when large-volume 

urine sample is necessary, such as for tritium or uranium determination.

Designated sample date is date of delivery to employee.

7, V Urine Approximate 12-hr collection Sunday-Monday sample (Friday delivery 

only). Generally used for workers chronically exposed to soluble uranium.

Designated sample date is Sunday in sampling period.

8, W Urine before 

1986. Feces 

starting in 

1986. 

Associated with urine sampling in 1950s through 1970s; used to mean 

undesignated or unknown. Starting in 1986: Collection of single fecal 

voiding used for special program for plutonium oxide workers. Designated 

sample date for shift workers is Tuesday of long shift change, and for day 

workers is appropriate Sunday.

9, X Urine Kit designed for collection of urine outside local service area.

Transportation handled by private carrier. Generally used for termination 

samples not collected locally.

A, Y Urine Simulated 48-hr urine collection. Collected at home over 4-d period. Used 

for IPUL (low-level isotopic plutonium) sampling. Designated sample date 

is 2 d after kit delivery to employee.

B, Not applicable Urine 12-hr urine collection for termination sampling only. Collected at home 

overnight. Kit delivered in normal manner, but brought to designated onsite 

location by worker for pickup by contractor. Designated sample date is day 

after date of kit delivery to employee. Delivery only, no home pickup 

required.

a. Before about 1983, kit codes were called collection codes.

b. D/R = Delivery and Retrieval; P/U = Pick-Up only (the latter series of codes were not used before about 1990, but 

should have no impact on dose reconstruction).
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ATTACHMENT B 

CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS (continued)

Table B-5. In vivo body location codes. 

Computer code Body location

ABD Abdomen

CHT Chest result

CH1 Chest result

CH2 Chest result corrected by ultrasound measurement of chest wall thickness

HED Head

HND Hand

KNE Knee

LG1 Lung result (chest result corrected for skeleton burden interference)

LG2 Lung result (chest result corrected for skeleton and liver burden interference)

LV1 Liver

LV2 Liver result corrected for skeleton burden interference

LV3 Liver result corrected for skeleton and lung burden interference

LYM Lymph nodes

SKL Skull (head) – old code no longer in use

SK1 Total activity in skeleton based on head count

SK2 Skeleton result based on something other than head count

SPL Special

THX Thorax

THY Thyroid

TRT Throat – old code no longer in use

WBC Whole body

WND Wound

Table B-6. Excreta unit codes. 

Computer code Description of units

1 dpm/sample

2 dpm/volume analyzed 

3 μg/L until 07-01-82

μg/sample after 07-01-82

4 μg/g until 07-01-82

μg/sample after 07-01-82

5 μCi/sample

6 μCi/L

7 nCi

8 μCi

Table B-7. Excreta processing codes.

a

Processing code Description

R Routine processing

P Priority processing

X Expedite processing (added about 1985)

E Emergency processing

a. Used in conjunction with contract with commercial laboratory 

starting in 1965. Used to designate turnaround time and MDAs; that 

is, different processing codes had different MDAs.
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ATTACHMENT B 

CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS (continued)

Table B-8. Excreta laboratory codes. 

Code Analytical laboratory

IT IT Analytical Services, Richland

LA Los Alamos National Laboratory

OR Oak Ridge National Laboratory

PL PNL Analytical Chemistry Laboratory

QN Quanterra

RE REECo (Reynolds Electrical & Engineering Company, 

Nevada Test Site)

ST Severn Trent Laboratories, Richland

TA TMA/Norcal, Richmond, California

WH Westinghouse Hanford Company, 222-S Laboratory

Table B-9. Excreta no-sample codes. 

No-sample code Description

CN Kit not out. Sample kit not out at time of scheduled pickup.

CS Cancelled sample or analysis.

CT Sample lost due to bioassay analysis contract termination.

FA Failed analysis. Valid analytical result could not be obtained.

IS Insufficient sample. Sample provided by worker but volume 

insufficient to meet contractual requirements.

LC Lost container. Sample kit not retrieved.

ND Not delivered. Sample scheduled but kit never delivered.

NS No sample. Kit retrieved but no sample provided by worker.

Table B-10. In vivo invalid result codes. 

Codea Reason for no results

C External contamination other than radon detected on subject. Measurement invalid; no results 

obtained.

F Failure of equipment or faulty setup of equipment. Measurement invalid; no results obtained.

I Interference from localized activity in another part of subject’s body. Measurement invalid; no results 

obtained.

L Location of internal or external activity qualitatively determined by mapping, masking, or collimating.

Could include one or more measurement counts. Measurements are qualitative for identifying 

location of activity and do not yield quantifiable estimates of activity.

M Medically administered radioactivity interfered with measurement. Measurement invalid; no results 

obtained.

P Preliminary count, when followed by more quantitative record count. Used to indicate measurement 

taken, but not to record count.

R Radon interference from subject's clothing, hair, or skin. Measurement invalid; no results obtained.

S Subject's actions interrupted completion of count. Measurement invalid; no results obtained.

X Measurement invalid; no results obtained. Other no-result codes do not apply. See comment field for 

brief description.

a. The comment field might contain a brief explanation in addition to the codes listed.

Table B-11. INTERTRAC mode-of-intake codes. 

Code Mode of intake

ABS Absorption

ING Ingestion

INH Inhalation

NON None (no intake)

UNK Unknown

WND Wound
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 118 of 180

ATTACHMENT B 

CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS (continued)

Table B-12. INTERTRAC evaluation reason codes. 

Code Reason for evaluation

A Annual chronic intake evaluation

C Contractor-requested evaluation

H High routine bioassay evaluation

I Incident evaluation

N New hire measurement or previous employment record 

indicated exposure before Hanford employment

R Reevaluation

Table B-13. INTERTRAC source-of-intake codes. 

Code Source of intake

DHE Intake at DOE site while employed at Hanford

HAN Intake at Hanford

NHE Intake at non-DOE site while employed at Hanford

NOC Nonoccupational intake

PTH Intake occurred before Hanford employment

Table B-14. INTERTRAC miscellaneous codes. 

Code type Code and description

Intake confirmed Y = Yes

N = No

Nature of intake A = Acute

C = Chronic

Recorded dose Y = Yes

N = No

O = Undetermined; old evaluation assessing body burden 

rather than dose, or evaluation in process

Z = Recorded dose is 0 mrem

Source known Y = Yes

N = No

Type of evaluation P = Preliminary

F = Final

Table B-15. Special whole-body count resolution codes (RC) (used in 

1983 only). 

Code Description

A Investigation in progress

B Recent intake < or = to 1% maximum permissible annual dose

C Previous deposition

D Under investigation with additional examinations scheduled

E Investigation completed, see radiation exposure records

F Unresolved

G Deposition from previous non-Hanford employment

H Exposure received offsite by Hanford employee

I Activity derived from medical diagnostic or therapeutic procedure
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 119 of 180

 

ATTACHMENT C 

INTERNAL DOSIMETRY CO-EXPOSURE DATA

TABLE OF CONTENTS

SECTION TITLE PAGE

C.1 Purpose...................................................................................................................................123

C.2 Overview................................................................................................................................. 123

C.3 Data ........................................................................................................................................125

C.3.1 Selected Bioassay Data .............................................................................................. 125

C.3.1.1 Plutonium Urinalysis.....................................................................................126

C.3.1.2 Uranium Urinalysis .......................................................................................129

C.3.1.3 Strontium Urinalysis .....................................................................................130

C.3.1.4 Promethium Urinalysis..................................................................................131

C.3.1.5 Zinc-65 in Whole-Body Counts.....................................................................132

C.3.1.6 Sodium-24 in Whole-Body Counts ...............................................................133

C.3.1.7 Cesium-137 in Whole-Body Counts .............................................................133

C.3.2 Analysis.......................................................................................................................134

C.3.3 Tritium Dose Evaluation..............................................................................................134

C.4 Intake Modeling.......................................................................................................................137

C.4.1 Assumptions................................................................................................................137

C.4.2 Bioassay Fitting...........................................................................................................138

C.4.3 Radionuclides and Material Types ..............................................................................139

C.4.3.1 Plutonium .....................................................................................................139

C.4.3.1.1 Type M ......................................................................................139

C.4.3.1.2 Type S.......................................................................................140

C.4.3.1.3 Type SS.....................................................................................140

C.4.3.2 Uranium........................................................................................................140

C.4.3.2.1 Type F .......................................................................................140

C.4.3.2.2 Type M ......................................................................................141

C.4.3.2.3 Type S.......................................................................................141

C.4.3.3 Strontium-90................................................................................................. 142

C.4.3.4 Promethium-147...........................................................................................142

C.4.3.4.1 Type M ......................................................................................142

C.4.3.4.2 Type S.......................................................................................142

C.4.3.5 Zinc-65 .........................................................................................................142

C.4.3.5.1 Inhalation Intakes ......................................................................142

C.4.3.5.2 Ingestion Intakes .......................................................................143

C.4.3.6 Sodium-24 ....................................................................................................143

C.4.3.6.1 Inhalation Intakes ......................................................................143

C.4.3.6.2 Ingestion Intakes .......................................................................144

C.4.3.7 Cesium-137 ..................................................................................................144

C.5 Assignment of Intakes and Doses ..........................................................................................144

C.5.1 Plutonium ....................................................................................................................145

C.5.2 Uranium.......................................................................................................................146

C.5.3 Strontium-90................................................................................................................147

C.5.4 Promethium-147..........................................................................................................148

C.5.5 Zinc-65 ........................................................................................................................148
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

C.5.6 Sodium-24...................................................................................................................149

C.5.7 Cesium-137.................................................................................................................150

C.5.8 Additional Radionuclides.............................................................................................150

C.6 Statistical Summaries and Plots .............................................................................................151

LIST OF TABLES

TABLE TITLE PAGE

C-1 Plutonium reporting levels.......................................................................................................128

C-2 Time intervals for statistical analysis of plutonium urine samples...........................................128

C-3 Uranium reporting levels .........................................................................................................129

C-4 Time intervals for statistical analysis of uranium urine samples .............................................130

C-5 Strontium reporting levels .......................................................................................................130

C-6 Time intervals for statistical analysis of 90Sr urine samples ....................................................131

C-7 Promethium reporting levels ...................................................................................................132

C-8 Time intervals for statistical analysis of 147Pm urine samples .................................................132

C-9 Time intervals for statistical analysis of 65Zn whole-body counts............................................133

C-10 Time intervals for statistical analysis of 24Na whole-body counts ...........................................134

C-11 Time intervals for statistical analysis of 137Cs whole-body counts ..........................................134

C-12 Plutonium mixture usage ........................................................................................................138

C-13 Type M 239+240Pu intake periods and rates ..............................................................................139

C-14 Type S 239+240Pu intake period and rates.................................................................................140

C-15 Type SS 239+240Pu intake period and rates ..............................................................................140

C-16 Type F uranium intake periods and rates ...............................................................................141

C-17 Type M uranium intake periods and rates...............................................................................141

C-18 Type S uranium intake periods and rates ...............................................................................141

C-19 Type F 90Sr intake periods and rates ......................................................................................142

C-20 Type M 147Pm intake periods and rates ..................................................................................142

C-21 Type S 147Pm intake periods and rates ...................................................................................142

C-22 Type S 65Zn inhalation intake periods and rates .....................................................................143

C-23 Zinc-65 ingestion intake periods and rates .............................................................................143

C-24 Type F 24Na inhalation intake periods and rates.....................................................................143

C-25 Sodium-24 ingestion intake periods and rates........................................................................144

C-26 Type F 137Cs intake periods and rates ....................................................................................144

C-27 Type M 239Pu intake periods and rates ...................................................................................146

C-28 Type S 239Pu intake period and rates......................................................................................146

C-29 Type SS 239Pu intake period and rates ...................................................................................146

C-30 Uranium type F intake periods and rates ................................................................................146

C-31 Uranium type M intake periods and rates ...............................................................................146

C-32 Uranium type S intake periods and rates................................................................................147

C-33 Type F 90Sr intake periods and rates ......................................................................................147

C-34 Type M 147Pm intake periods and rates ..................................................................................148

C-35 Type S 147Pm intake periods and rates ...................................................................................148

C-36 Zinc-65 inhalation intake periods and rates ............................................................................149

C-37 Zinc-65 ingestion intake periods and rates .............................................................................149

C-38 Type F 24Na inhalation intake periods and rates.....................................................................150

C-39 Type F 24Na ingestion intake periods and rates......................................................................150
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

C-40 Type F 137Cs intake periods and rates ....................................................................................150

C-41 Statistical summary of 239Pu 24-hour urinary excretion rates, 1946 to 1988...........................151

C-42 Statistical summary of uranium 24-hour urinary excretion rates, 1948 to 1988......................160

C-43 Statistical summary of 90Sr 24-hour urinary excretion rates, 1966 to 1988.............................166

C-44 Statistical summary of 147Pm 24-hour urinary excretion rates, 1967 to 1977..........................167

C-45 Statistical summary of 65Zn measured in whole-body counts, 1960 to 1984 ..........................169

C-46 Statistical summary of 24Na measured in whole-body counts, 1960 to 1984..........................174

C-47 Statistical summary of 137Cs measured in whole-body counts, 1960 to 1988.........................177

C-48 Hanford monthly reports .........................................................................................................179

C-49 Tritium hard-copy bioassay data.............................................................................................179

LIST OF FIGURES

FIGURE TITLE PAGE

C-1 Plutonium bioassay data in REX and from Hanford monthly reports......................................124

C-2 Uranium bioassay data in REX and from Hanford monthly reports ........................................125

C-3 Tritium bioassay data..............................................................................................................135

C-4 Individual tritium bioassay or dose data by year.....................................................................136

C-5 Tritium dose from hard-copy records and the main text .........................................................137

C-6 Production rates in tons of uranium times fuel burnup, 1944 to 1972.....................................145

C-7 50th-percentile plutonium urinalysis data for intakes of type M material, 1946 to 1948 .........154

C-8 50th-percentile plutonium urinalysis data for intakes of type M material, 1949 to 1952 .........155

C-9 50th-percentile plutonium urinalysis data for intakes of type M material, 1953 to 1981 .........155

C-10 50th-percentile plutonium urinalysis data for intakes of type M material, 1982 to 1988 .........155

C-11 Predicted 50th-percentile urinary excretion of type M plutonium from 1946 to 1988 based 

on four independent intakes, compared to bioassay data ......................................................156

C-12 50th-percentile plutonium urinalysis data for intakes of type S material, 1946 to 1948 ..........156

C-13 50th-percentile plutonium urinalysis data for intakes of type S material, 1949 to 1952 ..........156

C-14 50th-percentile plutonium urinalysis data for intakes of type S material, 1953 to 1981 ..........157

C-15 50th-percentile plutonium urinalysis data for intakes of type S material, 1982 to 1988 ..........157

C-16 Predicted 50th-percentile urinary excretion of type S plutonium from 1946 to 1988 based 

on four independent intakes, compared to bioassay data ......................................................157

C-17 50th-percentile plutonium urinalysis data for intakes of type SS material, 1946 to 1948........158

C-18 50th-percentile plutonium urinalysis data for intakes of type SS material, 1949 to 1952........158

C-19 50th-percentile plutonium urinalysis data for intakes of type SS material, 1953 to 1981........159

C-20 50th-percentile plutonium urinalysis data for intakes of type SS material, 1982 to 1988........159

C-21 Predicted 50th-percentile urinary excretion of type SS plutonium from 1946 to 1988 

based on four independent intakes, compared to bioassay data............................................160

C-22 50th-percentile uranium urinalysis data for intakes of type F material, 1948 to 1988.............163

C-23 50th-percentile uranium urinalysis data for intakes of type M material, 1948 to 1988 ............163

C-24 50th-percentile uranium urinalysis data for intakes of type S material, 1948 to 1952.............163

C-25 50th-percentile uranium urinalysis data for intakes of type S material, 1953 to 1956.............164

C-26 50th-percentile uranium urinalysis data for intakes of type S material, 1957 to 1961.............164

C-27 50th-percentile uranium urinalysis data for intakes of type S material, 1962 to 1974.............164

C-28 50th-percentile uranium urinalysis data for intakes of type S material, 1975 to 1983.............165

C-29 50th-percentile uranium urinalysis data for intakes of type S material, 1984 to 1988.............165

C-30 Predicted 50th-percentile urinary excretion of type S uranium from 1946 to 1988 based 

on six independent intakes, compared to bioassay data ........................................................165
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

C-31 50th-percentile strontium urinalysis data for intakes of type F material, 1965 to 1988 ...........167

C-32 50th-percentile promethium urinalysis data for intakes of type M material, 1966 to 1969 ......167

C-33 50th-percentile promethium urinalysis data for intakes of type M material, 1970 to 1979 ......168

C-34 Predicted 50th-percentile urinary excretion of type M promethium from 1966 to 1979 

based on two independent intakes, compared to bioassay data ............................................168

C-35 50th-percentile promethium urinalysis data for intakes of type S material, 1966 to 1969.......168

C-36 50th-percentile promethium urinalysis data for intakes of type S material, 1970 to 1979.......169

C-37 Predicted 50th-percentile urinary excretion of type S promethium from 1966 to 1979 

based on two independent intakes, compared to bioassay data ............................................169

C-38 50th-percentile zinc whole-body counting data for inhalation intakes of type S material, 

1960 to 1963...........................................................................................................................171

C-39 50th-percentile zinc whole-body counting data for inhalation intakes of type S material, 

1964 to 1967...........................................................................................................................172

C-40 50th-percentile zinc whole-body counting data for inhalation intakes of type S material, 

1968 to 1970...........................................................................................................................172

C-41 50th-percentile zinc whole-body counting data for inhalation intakes of type S material, 

1971 to 1972...........................................................................................................................172

C-42 50th-percentile zinc whole-body counting data for inhalation intakes of type S material, 

1973 to 1977...........................................................................................................................173

C-43 50th-percentile zinc whole-body counting data for inhalation intakes of type S material, 

1978 to 1984...........................................................................................................................173

C-44 Predicted 50th-percentile urinary excretion of inhaled type S zinc from 1960 to 1984 

based on six independent chronic intakes and one acute intake, compared to bioassay 

data.........................................................................................................................................173

C-45 50th-percentile zinc whole-body counting data for ingestion intakes, 1960 to 1984...............174

C-46 50th-percentile sodium whole-body counting data for inhalation intakes of type F 

material, 1960 to 1984 ............................................................................................................176

C-47 50th-percentile sodium whole-body counting data for ingestion intakes, 1960 to 1984..........176

C-48 50th-percentile 137Cs whole-body counting data for inhalation intakes of type F material, 

1960 to 1988...........................................................................................................................179
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

C.1 PURPOSE

There are instances of energy employees who, for a variety of reasons, were not monitored for 

internal exposure during the course of their employment at a DOE facility. In addition, there are 

incidents in which an employee’s monitoring records are incomplete or unavailable. In such cases, 

data from coworkers can be used to approximate an individual’s possible exposure. The purpose of 

this attachment is to provide the details of the calculation and assignment of intakes based on coexposure data from the Hanford Site (including PNNL) for the purpose of estimating unmonitored 

exposures. Many of the periods addressed by this co-exposure analysis include those covered by the 

groups of Hanford workers that have been included in the SEC. The information provided in this 

attachment for those periods can be used when dose reconstruction is required outside the scope of 

the SEC inclusions.

C.2 OVERVIEW

Analysis of Coworker Bioassay Data for Internal Dose Assignment, ORAUT-OTIB-0019 [ORAUT 

2005], describes the general process for analyzing bioassay data for the assignment of doses based 

on co-exposure results.

Bioassay results for Hanford were obtained as a copy of the REX database sent to the NIOSH HERB 

before the start of the dose reconstruction project. The REX data necessary to generate the bioassay 

statistics for the co-exposure data project were converted to another database program. Data in the 

converted tables were spot-checked against REX at Hanford, and no discrepancies were found.

To determine the completeness of the REX database, it was compared with statistics from monthly 

reports that were prepared by Hanford for 1946 through 1959 (where records were available). REX 

was found to contain data for over 90% of the plutonium bioassay samples that were indicated as 

collected in the monthly reports. For uranium, REX contained data for approximately 77% of the 

samples. Figures C-1 and C-2 depict the bioassay sample data available from each source for 

plutonium and uranium, respectively. Table C-46 contains the SRDB reference numbers for the 

monthly reports that were used for this comparison.

As is typical of a site with large operating reactors and spent fuel dissolution operations, there was 

potential at Hanford for exposure to many radionuclides including fission and activation products, 

plutonium, uranium, and tritium. In addition, Hanford handled thorium for part of its history. With the 

advent of in vivo counting, detection of a wide range of gamma-emitting fission and activation 

products became possible. In addition, bioassay monitoring of workers on small-scale or short-term 

projects for unusual radionuclides was performed at different times. Therefore, there are a number of 

different radionuclides in the excreta and in vivo data. For most of the radionuclides in the database, 

there were too few measurements to allow reliable statistical evaluation. Exposure to such 

radionuclides was arguably rare or limited to only a few workers on a specific project and would not be 

appropriate for this general co-exposure study.

The urinalysis data that were chosen for this co-exposure study because of general applicability and 

number of measurements were from plutonium, uranium, radiostrontium, and promethium analyses, 

although the 147Pm intakes have limited applicability (see Sections C.3.1.4 and C.5.4). Before 

whole-body counting became routine in 1960, bioassay monitoring for intakes of some fission 

products was performed using a radiochemical procedure called “fission product urinalysis.”
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Figure C-1. Plutonium bioassay data in REX and from Hanford monthly reports. 

Plutonium bioassay samples

0

300

600

900

1200

1500

1800

Jan-45 Jan-47 Jan-49 Jan-51 Jan-53 Jan-55 Jan-57 Jan-59

Date

#

REX Pu Monthly Reports Pu

This procedure involved chemical separation and counting (gross beta) of rare-earth fission products 

and radioisotopes of strontium. However, the results are difficult to interpret because the actual 

radionuclide or composition of the mixture of radionuclides in the urine was not determined and 

different radionuclides had different chemical yields and detection efficiencies. Therefore, intakes of 

24Na, 65Zn, 90Sr, and 137Cs were extrapolated from periods when they were measured, and other 

fission products were accounted for using air concentration limits (see Section C.5). 

With the exception of naturally-occurring 40K, only three gamma-emitting radionuclides measured by 

whole-body counting were routinely reported: 137Cs, 24Na, and 65Zn. The available documentation 

does not state the reasons for choosing these three radionuclides. Other fission or activation products 

were reported only when detected above certain levels of significance that were determined through 

different methods at different times. A review of the in vivo database showed that there was enough 

data to warrant statistical evaluation only for these three routinely reported radionuclides.

Dose from intakes of tritium was treated as external dose until about 1986 and reported as part of 

penetrating dose in the radiological records. Dose from tritium was included in the co-exposure

analysis of penetrating external dose and was therefore not duplicated in this co-exposure analysis. 

The completeness of those records was subsequently evaluated as discussed in Section C.3.3. (The 
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Figure C-2. Uranium bioassay data in REX and from Hanford monthly reports. 

Uranium bioassay samples

0

100

200

300

400

500

600

Jan-45 Jan-47 Jan-49 Jan-51 Jan-53 Jan-55 Jan-57 Jan-59

Date

#

REX U Mon Rep U

co-exposure external analysis is in Attachment B of ORAUT-TKBS-0006-6, Hanford Site – 

Occupational External Dose [ORAUT 2010b].) 

The statistical analyses of the bioassay data for each radionuclide were performed in accordance with 

ORAUT [2005b]. The resultant values were input to the IMBA or Internal Dosimetry Tool (IDOT) 

computer program, and a fit to the data for each of the seven radionuclides at the 50th- and 

84th-percentile values was performed to obtain intake rates for assignment of dose distributions.

C.3 DATA

C.3.1 Selected Bioassay Data

Data for each of the database codes that were considered were extracted from a series of files that 

contained a version of the HERB dataset.

Throughout most of Hanford’s history, bioassay data were truncated or censored below some level.

Various terms can be applied to such a level, such as detection limit, less-than level, reporting level, 

lower limit of detection, decision level, or MDA. In this attachment, reporting level denotes the level 
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below which a measurement was not recorded. Decision level and MDA, as best as could be 

determined from available documentation, are used in accordance with the definitions in Health 

Physics Society Standard N13.30 [ANSI/HPS 1996]. The analysis established time intervals for which 

single values for the 50th- and 84th-percentile values were calculated. These intervals were 

determined until there was a sufficient total number of measurements and a sufficient number of 

measurements greater than the reporting level to perform meaningful statistical analysis.

For some data, reporting levels were not used for all samples, resulting in a large batch of samples at 

or just above the reporting level, a smattering of samples below the reporting level, and some 

samples with zero, blank, or null values. This mix of reporting practices led to poor curve fits when 

values at the reporting level were treated as zeros. Therefore, this analysis used the linear distribution 

[ORAUT 2005b] for values considered to represent nondetections. The following radionuclide-specific 

sections discuss exceptions to this treatment.

C.3.1.1 Plutonium Urinalysis

Routine urinalysis for plutonium at Hanford began in 1946. Over the years, several factors in the 

methods for measuring and reporting plutonium urinalysis data changed:

• Chemical method,

• Radioactivity counting method,

• Reporting units,

• Reporting levels and data censoring (that is, not reported below a certain activity),

• Method for indicating data below the reporting level, and

• Consistency of use of the previous two protocols (the data for some years show that the 

reporting level was not used consistently and/or that different methods of reporting data below 

the reporting level were used).

Some of these changes affected how the urinalysis data were treated for the statistical analysis. The 

issues and resolutions are discussed in the following paragraphs.

From 1946 to the third quarter of 1983, plutonium urinalysis results were measured and recorded as a 

gross alpha count on a chemically separated sample. The measured result was the total activity of 

238Pu, 239Pu, and 240Pu but did not include 241Pu or 241Am. Since the fourth quarter of 1983, the

chemically separated sample was measured by alpha spectrometry (for routine and priority samples).

The recorded results were a value for 239Pu that was actually the total activity of 239Pu and 240Pu and a 

value for the 238Pu activity. The statistical analyses were conducted on the total plutonium alpha 

activity for 1946 to September 1983 and on the 239+240Pu activity for October 1983 to 1988. Because 

most plutonium exposures at Hanford would have involved a mixture of isotopes, the results of these 

different measurement techniques were normalized (Section C.4). 

Routine plutonium urinalysis samples were usually simulated 24-hour samples that were collected 

over two 12-hour periods (usually consecutive evenings). Some samples were not 24-hour samples, 

especially those collected shortly after a suspected acute intake. The collected volume of urine was 

part of the record starting irregularly in 1958 and consistently in 1959, so the 1959 to 1988 results with 
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volumes less than 400 mL were not used in the statistics. This criterion removed some of the incident 

samples and some samples from persons who did not follow instructions and provided less than 

simulated 24-hour samples.

Incident samples were problematic for performing group statistical analysis. Hanford generally 

collected many samples from a worker involved with a known large acute intake, which were referred 

to as special or followup samples. The number of samples varied dependent on the severity of the 

intake, whether treatment was involved, the willingness of the exposed worker, and general internal 

dosimetry practices at the time. However, in many cases the number of followup samples ranged from 

tens to hundreds. These samples were not comparable to routine annual or semiannual samples from 

the majority of the workforce and they tended to bias the statistics on the high side. It was too difficult 

to remove all of these samples from the database, and minor acute intakes were frequent enough at 

Hanford that they could represent the normal course of exposure to plutonium workers. However, 

samples were removed if: 

• There were a large number of samples from a single worker, usually bunched after a specific 

time. A specific number was not rigorously applied as the definition of large, but it was 

generally greater than about 20 in 1 year. In addition, once a person was selected to be 

removed, data for subsequent years in that worker’s history were removed until the excretion 

was near or less than the reporting level.

• A record of an acute intake was judged to be unrepresentative. Any intake that was treated by 

diethylene triamine pentaacetic acid (DTPA) met this criteria because DTPA-treated urinary 

excretion is enhanced. Intake through a wound or the skin was considered unrepresentative. 

Inhalations, however, were considered representative unless they were caused by an event 

that was judged to be unrepresentative, such as explosion, fire, or major contamination 

spread.

Another challenge to data interpretation was results listed as “PU” or “PU239” with blank quantities.

For most years, the meaning of a blank was not found in documentation. However, the meaning could 

be inferred from the number of blank results and how they fit with the other data for the year. For 

instance, there were many blank results each year in the early years, and it was inferred that the 

blanks meant a sample was analyzed and nothing was detected. In later years many sample results 

were recorded exactly at the recording level with only a few blanks. It was favorable to claimants to 

leave out the blanks in those years under the assumption that a blank was a sample collected for 

which the analysis failed in some way. Each year’s distribution of sample results was scrutinized, and 

instructions were written on how to disposition blank results on a year-by-year basis.

The method that was used to indicate no detection of plutonium, similar to the issue of blanks, varied 

over time and appeared not to be consistent even during a given period. For instance, zero, blank, 

and the reporting level were apparently used to indicate nondetections during some periods. In 

general, Hanford used reporting levels for plutonium urinalysis results until September 1981. Whether 

a value that was recorded at the reporting level meant no detection, or whether the reporting level or 

higher was a true detection and nondetections were recorded as zero or blank, was indeterminate at 

the broad level.

The meanings of the reporting level, zero, and blanks had to be determined on a year-by-year basis, 

and the decision was included in the specific instructions on how to rank such sample results. For 

example, the 1966 to 1974 reporting level was 1.1 × 10-8 μCi/sample (0.244 dpm/sample) and the 

1975 to 1981 reporting level was 0.025 dpm/sample. These numbers were considered to be one-half 
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of the detection level and were used to indicate no detection of plutonium. The linear distribution was 

used to distribute and rank blanks, zeros, and values at these reporting levels. For other years, 

however, the reporting level was less apparent because there were many values that were probably a 

reporting level, but other values were less than the reporting level. The latter samples were included 

in the linear fit, and the apparent reporting level was set as the top of the linear distribution. Table C-1 

lists these reporting levels for various periods.

Table C-1. Plutonium reporting levels.

Period Reporting level

1946 3.6E-7 μCi/sample

1947–1948 3.0E-7 μCi/sample

1949–1951 1.5E-7 μCi/sample

1952 8.1E-8 μCi/sample

1953–1957 2.3E-8 μCi/sample

1958–1965 1.2E-8 μCi/sample

1966–1974 1.1E-8 μCi/sample

1975–September 9, 1981 1.1E-8 μCi/sample

(0.025 dpm/sample)

On September 10, 1981, the recording practice changed from recording 0.025 dpm/sample to indicate 

a nondetection to recording the exact result as measured. After September 9, 1981, any result greater 

than zero was used as recorded. Zero and negative results were included in the ranking of samples 

but not in the fitting of the line. The linear distribution was not used for data after September 9, 1981.

Units for plutonium analyses in the Hanford records were microcuries per sample from 1946 to 1974 

and disintegrations per minute per sample from 1975 to the present. The earlier values were 

converted to disintegrations per minute before statistical analysis.

The time interval for separate statistical analyses varied. The choice of the interval was based on the 

number of results versus the number of results over the reporting level. If there were several hundred 

sample results and 10% or more exceeded the reporting level, the basic time interval was quarterly.

Table C-2 lists the intervals. The unusual division of the third and fourth quarters of 1981 was applied 

because of the change in recording practice that is discussed above.

Table C-2. Time intervals for statistical analysis 

of plutonium urine samples. 

Calendar year Analysis interval

1946 Year

1947 Year

1948 Half years

1949 Year

1950 Year

1951 Year

1952 Year

1953 Half years

1954–June 1981 Quarter years

3Q 1981 July 1–September 9

4Q 1981 September 10–December 31

1982–1988 Quarter years
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C.3.1.2 Uranium Urinalysis

Routine urinalysis for uranium began in 1947, but the results were considered unreliable until 

improvements in the procedure were made [Healy 1948]. The 1947 data were analyzed but not used 

for intake modeling. Data analyzed for this study were from elemental analysis procedures. Starting in 

1983, an alpha spectrometry procedure has been used for some workers, but these data were less 

robust and were not representative of the overall Hanford workforce. The results were recorded as 

micrograms per liter from 1947 to July 1, 1982, and as micrograms per sample from July 2, 1982, to 

the present. The latter were converted to micrograms per liter using the sample volume that was 

recorded as part of the information in the database [130]. Elemental uranium samples were not 

usually 24-hour samples; a mix of sample collection periods was used from 1947 to 1988, and

overnight sampling was one of the more frequent methods. Therefore, samples of less than 400 mL 

were not removed from the uranium data.

In general, Hanford did not have many large acute intakes of uranium; only six samples were 

removed because they were collected after unusual acute intakes (e.g., a worker fell into a vat of 

uranyl nitrate) [131].

A small number of duplicate entries were found in the records from 1965 and 1968 that were true 

duplicates rather than, for instance, two analyses of the same sample; one of each pair was removed 

from the data before statistical analysis.

Table C-3 lists the reporting levels.

Table C-3. Uranium reporting levels.

Period Reporting level

1948–1974 4.0 μg/L

1975–1981 0.4 μg/L

1982–present Actual values were recorded as measured

Use of the reporting levels for uranium urinalysis appears to have been more consistent than was the 

case for plutonium, with the exception of one period. From 1958 to 1966, many results were lower 

than 4 μg/L. It is not known if these indicated detection (i.e., whether the MDA was really reduced 

during this period). Therefore, these were treated as less-than results and included in the linear 

distribution. In 1967, Hanford returned to consistent use of the 4-μg/L reporting level.

Through 1981, zeros, blanks, or less-than results were treated as nondetections and included in the 

linear distribution. Starting in 1982:

• No reporting level was used.

• The linear distribution was not used in the statistical analysis.

• Blanks were considered invalid samples and not used in the statistical analysis.

• Zero and negative numbers were included in the ranking of samples but not in the fitting of the 

line.

Table C-4 lists the time intervals for separate statistical analyses. The PUREX and UO3 plants were 

shut down during the 1970s, so there were few exposures and few samples [ORAUT 2010a]. 

Only 65 samples were collected in 1973; the results were comparable to neighboring years, so the 

1973 statistical parameters were included.
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Table C-4. Time intervals for statistical 

analysis of uranium urine samples. 

Calendar year Analysis interval

1947 Year

1948 Half years

1949–1969 Quarter years

1970–1983 Year

1984–1988 Quarter years

C.3.1.3 Strontium Urinalysis

Specific urinalysis for radiostrontium began in 1965. Before that, the fission product urinalysis method 

was used to monitor for radiostrontium and some other fission product radionuclides. Both 89Sr and 

90Sr could have been counted depending on whether the sample was beta-counted before or after a 

period for ingrowth of 90Y. For this study, all results were assumed to be 90Sr because it was generally 

more prevalent and because it produces higher (more favorable to claimants) doses per unit intake 

than 89Sr. Results were recorded as microcuries per liter for 1965 to 1974 and as disintegrations per 

minute per sample from 1975 to the present. The earlier results were converted to disintegrations per 

minute per sample before statistical analysis using the recorded volume of the sample.

Most strontium urine samples were simulated 24-hour samples, so samples that were less than 

400 mL were removed from the data before statistical analysis. Thirteen sample results were removed 

because they were followup samples to known large acute intakes; one sample result was removed 

because it was a recount. There were some obvious duplicates that were removed.

Table C-5 lists the reporting levels. 

Table C-5. Strontium reporting levels.

Period Reporting level

1965–1969 37 dpm/L (1.67E-5 μCi/L)

1970–1974 2.2 dpm/L (1.00E-6 μCi/L)

1975–03/1979 2 dpm/sample

04/1979–1981 5 dpm/sample

1982–1988 Actual values were recorded as measured

Blanks with volumes greater than 400 mL were rare. These were considered invalid and were not 

used. Through 1981, zero or less-than results were treated as nondetections and included in the 

linear distribution. Starting in 1982:

• No reporting level was used.

• The linear distribution was not used in this analysis.

• Zero and negative numbers were included in the ranking of samples but not in the fitting of the 

line.

Table C-6 lists the time intervals for separate statistical analyses. Before 1970, there were few 

samples in a given year and very few samples that exceeded the reporting level, so samples were 

grouped into two 3-year intervals. Before 1979, in general, there were only a few hundred (and

sometimes less than a hundred) samples in a year.
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Table C-6. Time intervals for statistical analysis of 90Sr 

urine samples. 

Calendar year Analysis interval

1965, 1966, 1967 As one group

1968, 1969, 1970 As one group

1971–1978 Years

1Q 1979 Quarter

2,3,4Q 1979 As one group

1980–1981 Years

1982–1988 Quarters

Whole-body counting was the primary bioassay method for fission products; the exception was for the 

workers at the Waste Encapsulation and Storage Facility where exposure to pure 90Sr was possible.

The data from 1979 were split into two periods because of the change in reporting level from 2 to 

5 dpm/sample on April 1.

C.3.1.4 Promethium Urinalysis

Promethium-147 is a fission product with greater abundance in 1-year-aged weapons-grade fuel 

(1 year since removal from the reactor) than either 90Sr or 137Cs. However, it behaves in the body 

somewhat similarly to 90Sr and had an early MPBB that was a factor of 30 less than that for 90Sr [ICRP 

1959] and, later, an annual limit on intake that was a factor of 14 less than that for 90Sr [ICRP 1979, 

1982]. The presence of 147Pm would have been detected as part of the fission product urinalysis even 

though the results of the fission product urinalysis were calibrated for and interpreted as 90Sr. When 

whole-body counting and 90Sr urinalysis replaced the fission product urinalysis as the bioassay 

method of choice for fission products, 147Pm was considered an insignificant contributor to internal 

dose and 147Pm bioassay for workers exposed to general fission products was not performed.

Promethium-147 as an unconcentrated fission product in general contamination was treated in this 

analysis as a miscellaneous fission product (see Section C.5.8). 

Specific bioassay for 147Pm was initiated when PNNL began manufacturing 147Pm heat sources in the 

325 Building in 1966. The number of workers in the bioassay program and the number of samples 

were small in comparison to the numbers for the plutonium, uranium, or strontium bioassays, with a 

high of 65 workers in 1968. There was almost no sampling from 1972 to 1975 – which likely indicates 

a cessation of the original heat source program – and only 20 workers were sampled from 1976 to 

1979. The exact end date of the heat source program has not been determined. For purposes of this 

study, it was assumed that exposure occurred from 1966 to 1979, although it is likely that any 

exposure from 1972 to 1975 was only due to residual activity.

Results were recorded as microcuries per liter from 1966 to 1974 and as disintegrations per minute 

per sample from 1975 to 1979. The earlier results were converted to disintegrations per minute per 

sample using the sample volume.

Most routine samples were simulated 24-hour samples, so samples with volumes smaller than 

400 mL were removed from the data before statistical analysis. A large number of duplicate entries in 

the database from 1966 to 1968 were also removed. There was a glovebox rupture and large 

contamination spread that mainly affected three workers in 1967, and an overpressurization and 

blowback of a sampling line affected another worker in 1968; samples that were associated with these 

incidents were removed [132].

Table C-7 lists the reporting levels.
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Table C-7. Promethium reporting levels.

Period Reporting level

1966–03/25/1970 37 dpm/L (1.67E-5 μCi/L)

03/31/1970–1974 22 dpm/L (1.00E-5 μCi/L )

1975–1979 25 dpm/sample

The linear distribution was applied to results at or below the reporting levels. There were no blanks 

with volumes greater than 400 mL.

Table C-8 lists the time intervals for separate statistical analyses. The split in 1970 occurred because 

of the change in reporting level on March 31.

Table C-8. Time intervals for statistical analysis of 147Pm urine samples. 

Calendar year Analysis interval

1966, 1967 As one group

1968, 1969, 01/01/1970–03/25/1970 As one group

03/31/1970–12/31/1970; 1971 As one group

1972–1974 Not enough samples for statistics

1975, 1976, 1977, 1978, 1979 As one group

C.3.1.5 Zinc-65 in Whole-Body Counts

Zinc-65 was routinely reported in whole-body counts from 1960 to 1983. There were only about 

370 recordings of 65Zn in whole-body counts in 1984, but that was considered sufficient to include in 

the statistical analysis. After 1984, there were very few detections of 65Zn in whole-body counts.

Zinc-65 was mostly a concern during the operation of the once-through-cooled reactors (through 

1971) when large activities of 65Zn were discharged to the Columbia River.

The three largest local cities drew their sanitary water mostly from the Columbia River, as did the 

reactors on the site. Brady [1964] shows that 65Zn passed through the purification systems in sufficient 

concentration to produce measurable body burdens in many Hanford workers. Brady also attributed 

some of the 65Zn to ingestion of fish from the Columbia River or from crops irrigated by Columbia 

River water.

Figure C-35 shows a decrease in the median body burden of 65Zn starting in 1972 (with the exception 

of a spike in body burden in the fourth quarter of 1976), which appears to show the effect of the 

shutdown of the once-through-cooled reactors. However, it was apparent from the distribution of 

measured activities in the workers that inhalation in the workplace also had to be a viable intake 

pathway; ingestion of water cannot explain the higher body burdens at the higher percentile values of 

the quarterly distributions. Therefore, both inhalation and ingestion pathways were modeled.

In whole-body counting, the decision level varies with every individual and every count because of the 

interperson variability in activity of 40K, the intraperson variability in activity of interfering radionuclides 

a person could have on them on any given day (especially radon and thoron progeny), and the 

day-to-day variability in the natural background in the counting cell. In addition, and especially during 

the early years of the in vivo counting program, there were improvements to the program including 

type or number of detectors, counting times, and improved electronics. It was common for more than 

one whole-body counting system to be used at a time. Therefore, the stated detection limits were 

approximate and were meant to apply to a general range; little documentation on detection levels has 

been found. A single reporting level is not apparent in the recorded data. All whole-body counts were 

recorded in nanocuries.
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Because of these factors, it was impossible to state a distinct threshold value above which all results 

meant detection and below which all results meant nondetection. For 1960 to 1974, a detection level 

of 0.8 nCi for 65Zn was found in the documentation. Relative to the distribution of results in most 

years, this value was a reasonable one to use as the reporting level. Blanks and zero results were 

distributed using the linear distribution with 0.8 nCi as the upper range; values between zero and 

0.79 nCi were used as recorded and mixed in with the numbers from the distribution.

For 1975 to 1984, the value of 0.75 nCi was distinctly used as the reporting level, and only one value 

fell between zero and 0.75 nCi. Blanks and zeros were distributed using the linear distribution.

Table C-9 lists the time intervals for separate statistical analyses.

Table C-9. Time intervals for statistical 

analysis of 65Zn whole-body counts. 

Calendar year Analysis interval

1960 Year

1961–1962 Half years

1963–1974 Quarter years

1975 Half years

1976–1983 Quarter years

1984 Year

C.3.1.6 Sodium-24 in Whole-Body Counts

Similar to 65Zn, 24Na was reported routinely from 1960 to 1983 with a significant but much smaller 

number of results in 1984; very few results were recorded after 1984. Sodium-24 was released to the 

Columbia River from the once-through-cooled reactors, but its short half-life (15 hours) prevented 

significant concentrations in local municipal drinking water systems. Its presence was correlated with 

reactor workers and was to some extent believed to be due to drinking water at the reactors [Brady 

1964]. However, as with 65Zn there were body burdens that were too high to be explained by ingestion 

of drinking water. As a consequence, 24Na was modeled using both inhalation and ingestion 

pathways.

As explained in Section C.3.1.5 for 65Zn, a distinct threshold value for detection was not apparent from 

the data. For 1960, the lowest reported value was 0.56 nCi, so this value was used as the reporting 

level. Blanks and zero results were distributed using the linear distribution with 0.56 nCi as the upper 

range. For 1961 to 1974, a detection level of 0.3 nCi for 24Na was found in the documentation. This 

value was reasonable for use as the reporting level even though there were many results recorded at 

smaller values. Blanks and zero results were distributed using the linear distribution with 0.3 nCi as 

the upper range. Values between zero and 0.29 nCi were used as recorded and mixed with the 

numbers from the distribution. For 1975 to 1984, a value of 0.31 nCi was distinctly being used as the 

reporting level and no measurements lower than 0.31 nCi were recorded. Blanks and zeros were 

distributed using the distribution with 0.3 nCi as the upper range.

Table C-10 lists the time intervals for separate statistical analyses.

C.3.1.7 Cesium-137 in Whole-Body Counts

Cesium-137 was routinely recorded with every whole-body count from 1960 to 1983. For 1984 to 

1988, 137Cs was generally recorded only if it was detected above a reporting level, but the database 

also shows valid whole-body counts including blank values for 137Cs. Why this occurred is not known.
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Table C-10. Time intervals for statistical 

analysis of 24Na whole-body counts. 

Calendar year Analysis interval

1960 Year

1961–1962 Half years

1963–1983 Quarter years

1984 Half years

Because of its long half-life and abundance in the waste from the separations plants, 137Cs has long 

been considered one of the principal contaminants at Hanford and has often been used as an 

indicator of potential intakes. The linear distribution was not used to evaluate the 137Cs data except for 

the period of 1984 through 1988.

Table C-11 lists the time intervals for separate statistical analyses.

Table C-11. Time intervals for statistical analysis of 137Cs 

whole-body counts. 

Calendar year Analysis interval

1960–1974 Quarter years

1975 Only two values reported and neither above 

the reporting level. This year was skipped.

1976–1983 Quarter years

1984–1988 Half years

From 1984 to June 1986, 137Cs results were recorded differently than in other years. When 137Cs was 

not detected, the radionuclide was sometimes listed in the results as a blank but, more often, 137Cs 

was not listed at all. Instead, it was treated like the many other fission product radionuclides that could 

be detected by in vivo counting but are not listed if they are not detected. For this period, the total 

number of whole-body counts was determined by tallying the number of 40K counts, which were 

recorded for every valid whole-body count. Each whole-body count for which 137Cs was not listed was 

considered the same as a whole-body count with a blank or zero result for 137Cs.

The numbers of results recorded as greater than the reporting level (0.65 nCi for 1984 to 

September 19, 1986, and 3 nCi for September 22, 1986, to 1988) were very small (generally less than 

2% of the total counts during each half-year period). The rank-only method forces a fit to these small 

numbers of detected results and extrapolates the fits to the almost 4,000 counts in each half-year 

period that had no detection. Therefore, this analysis used the linear distribution for 1984 to 1988 

because it provides better fits (visually and by the R2 values).

C.3.2 Analysis

For each of the seven radionuclides, a lognormal distribution for the data in the intervals in 

Tables C-2, C-4, C-6, and C-8 to C-11 was assumed. The 50th- and 84th-percentile values were 

calculated using the method in ORAUT [2005]. Tables C-39 to C-43 show the statistical analysis 

results for plutonium, uranium, 90Sr, 147Pm, 65Zn, 24Na, and 137Cs.

C.3.3 Tritium Dose Evaluation

The tritium doses in the main text of this document are based on calculated summary doses in the 

Hanford external dosimetry records. However, bioassay data are typically preferred for dose 

reconstruction over summary doses that are calculated contemporaneously with the exposure
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because assumptions for this project are not necessarily consistent with those of a radiation 

protection program.

In 2007, several boxes of tritium data were identified that appeared to contain tritium urinalysis data 

from 108-B workers for the late 1940s and 1950s. Additional boxes were subsequently identified and 

expanded the period into the 1960s. In 2008, the contents of these boxes became available to the 

ORAU Team. The hard-copy records were entered into a database to permit analysis. The calculated 

doses from these data can be compared to the summary dose data that was previously the only 

available data. The objective of this comparison was to determine whether the tritium doses to be 

assigned to unmonitored workers as provided in the main text of this document (Section 5.6) are 

believed to be favorable to claimants. As a practical matter, the objective became whether there is 

any reason to believe that they are not favorable given the limitations of the comparisons that can be 

performed.

Figure C-3 summarizes the data. Table C-46 contains a list of the SRDB reference numbers for the 

monthly reports. Table C-47 contains a list of the SRDB reference numbers for the hard-copy tritium 

bioassay data.

Figure C-3. Tritium bioassay data. 

Tritium Bioassay Data

0

2000

4000

6000

8000

10000

12000

14000

1945 1950 1955 1960 1965 1970

Year

# samples

Hardcopy data Hardcopy data >MDA Monthly Reports

The number of bioassay samples in the hard-copy records follows the trend of the number of tritium 

bioassay samples that were reported as being taken in the Hanford monthly reports for most years for 

which a comparison could be made except 1961 through 1964. No bioassay data was found for 1961 

through 1963 and very little was found for 1964. There does appear to be a gap in the hard-copy 

records for 1961 through 1964 and 1966 where the number of workers with reported doses in the 

external dosimetry database exceeds the number for which hard-copy records were found. In 

addition, the monthly report data is known to be incomplete because the reports for all months in the 

period of interest are not available. Annual tritium bioassay totals or reports with tritium bioassay 

information for all months in a given year are available only for 1950 and 1963. The tritium bioassay in 
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the hardcopy records exceeded that reported in the available monthly reports for 1949 through 1952, 

1954, and 1967. However, the tritium bioassay in the hard-copy records are also known to be 

incomplete based on known instances where the retrieved bioassay data are less than that reported 

in the incomplete monthly reports, especially 1961 through 1964 and 1966. An additional complicating 

factor is that the monthly reports track analyzed bioassay samples rather than collected samples, 

which might introduce an offset into any comparison because the tritium bioassay in the hard-copy

records is recorded based on the sample collection date.

The number of individuals with recorded doses in the REX database was compared to the number of 

individuals for whom doses were calculated based on the hard-copy records. In addition, the doses 

were compared on an individual-specific basis where dose data for an individual was present in both 

the REX data and in the hard-copy records (Figure C-4). Relative success was achieved in matching 

individuals in both datasets. The number of individuals with data in both datasets closely tracked the 

minimum number of individuals in either of the data sources for a given year.

Figure C-4. Individual tritium bioassay or dose data by year. 

Individuals by Year

0

50

100

150

200

250

300

350

400

1945 1950 1955 1960 1965 1970

Year

# individuals

REX Hardcopy Matched

Using an updated version of a tool for calculation of co-exposure doses from tritium bioassay for the 

Mound and LANL co-exposure studies, tritium doses were calculated for each worker and year for 

which data was available. The doses by individual were compared on an annual basis to determine if 

a correlation between the main text (REX) values and the dose based on the hard-copy records could 

be determined. The most relevant comparison of the two data sources is the GM of the annual doses 

for the individuals in each dataset for each year because this is the actual dose to be used for 

unmonitored workers. The distributions of annual doses for all individuals in a given year were fitted to 

a lognormal distribution for each year. Figure C-5 depicts the comparison of the calculated GM doses 

from these two data sources. As can be seen, the values in the main text are higher than the values 

calculated from the hard-copy records in all but 1 year. The values in Table 5-42 are already 
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Figure C-5. Tritium dose from hard-copy records and the main text (Section 5.6). 

normalized to current-day tritium modeling. The unnormalized main text values are similar in 

magnitude and generally higher than the doses based on the hard-copy records as well.

The doses that would be assigned by either method are not significant at less than 200 mrem/yr in all 

years and generally much lower. Given that the values in the main text are favorable to the claimant in 

comparison with the values that were calculated from the hard-copy records, and that they cover more 

years than those values, the values assigned in the main text are deemed more appropriate for 

assignment of co-exposure doses.

C.4 INTAKE MODELING

C.4.1 Assumptions

All urinalysis results were assumed to be representative of a full-day (24-hour) urinary excretion or 

were either eliminated or normalized to be representative of a full day before creating the lognormal 

plots [133]. Each result that was used in the intake calculation was assumed to have a normal 

distribution, and a uniform absolute error of 1 was applied to all results, which weighted all results 

equally. A chronic exposure pattern was assumed even though this is unlikely for most exposures at 

Hanford because it approximates a series of acute intakes with unknown dates. Intakes were 

assumed to be from inhalation using a default breathing rate of 1.2 m3/hr and a 5-μm activity median 

aerodynamic diameter particle size distribution unless otherwise specified [NIOSH 2002]. Intakes of 

65Zn and 24Na were determined using whole-body counting results. As discussed in Sections C.3.1.5

and C.3.1.6, both inhalation and ingestion intakes were possible, so both were modeled.
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The results for 147Pm and 90Sr urinalysis were taken directly from the statistical analysis of the 

database, as were the results for 65Zn, 24Na, and 137Cs whole-body counting.

For plutonium, 239Pu was used for the intake modeling, but the database results were presumed to be 

gross plutonium alpha measurements through the third quarter of 1983. To convert from gross 

plutonium alpha to 239Pu, fresh fuel-grade plutonium was assumed for 1945 to 1949, 5-year-aged 

fuel-grade plutonium was assumed for 1950 to 1954, and 10-year-aged fuel-grade plutonium was 

assumed for 1955 and after [134]. Table C-12 lists the percentages of 239Pu that were used. Starting 

in October 1, 1983, alpha spectrometry was used in the analysis to specifically identify the 239Pu; 

therefore, no correction factor was used.

Table C-12. Plutonium mixture usage. 

Period Plutonium grade

Percentage Pu-239 of 

plutonium alpha emitters

09/01/1946–1949 Fresh fuel-grade 82.4%

1950–1954 5-yr-aged fuel-grade 83.0%

1955–09/30/1983 10-yr-aged fuel-grade 83.5%

10/01/1983–

present

10-yr-aged fuel-grade 100% (isotopic analysis)

For uranium, because the IMBA program requires urine results in units of activity per day, the total 

uranium values in micrograms per liter were multiplied by 1.4 to normalize them to the Reference Man 

excretion rate of 1,400 mL/d. Because a variety of enrichments was possible, the intake modeling 

used mass concentration units and 234U was assumed for all the IMBA intake modeling because it has 

the dose conversion factors most favorable to claimants [ICRP 1994] [135]. This did not affect the 

data fits for intake determination (i.e., the same total intakes would be obtained for any enrichment 

that was assumed) because all uranium isotopes have the same biokinetic behavior and the isotopes 

in this analysis have long half-lives in relation to the assumed intake period.

To convert to units of activity for dose calculation, the 0.7054 pCi/μg specific activity of NU was 

assumed through 1952, and the 0.9099 pCi/μg specific activity of RU was used for later years (see 

the main text). Like the intake modeling, 234U was used to calculate the doses. The ICRP 

Publication 68 dose coefficients (also referred to as dose conversion factors) for 234U are 7% to 31% 

larger than those for 235U, 236U, and 238U [ICRP 1994b]. Because of the isotopic compositions of the 

source terms, the 234U dose conversion factor yields doses that are favorable to claimants.

C.4.2 Bioassay Fitting

The IMBA and IDOT computer programs were used to fit the bioassay results to a series of chronic 

inhalation or ingestion intakes. The exception is that one acute intake was used in the modeling of 

24Na exposure because that resulted in the best fit. Modeling was performed only for intervals when 

data were available. Having been based on whole-body counts, the data for 65Zn and 24Na were 

available for 1960 to 1984. The least amount of available data was for 147Pm from 1966 to 1979.

Plutonium data was from 1946 to 1988, uranium data was from 1948 to 1988, and 90Sr data was from 

1965 to 1988.

The intake assumptions were based on observed patterns in the bioassay data. Intervals with 

constant chronic intake rates were chosen by selecting periods during which the bioassay results 

were of similar magnitudes. A new chronic intake period was started where the data indicated a 

significant sustained change in the results.
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C.4.3 Radionuclides and Material Types

For each considered radionuclide or group of radionuclides, the bioassay results were entered into 

IMBA or IDOT with assumed material types in terms of lung absorption type or uptake factor from the 

gastrointestinal tract. The types were chosen to be consistent with ICRP Publication 68 [ICRP 1994b] 

and the main text of this document. The figures at the end of this attachment show the resultant 

50th-percentile intakes as plots. The annual bioassay data that were used in the fits are shown as 

solid blue dots (●) (dark dots when printed in black and white) and data that were not used in the fits 

are shown as red dots (●) (light dots when printed in black and white).

The type S and super S (SS) compounds at Hanford have very long radiological half-lives and the 

materials are retained in the body for long periods. Therefore, the excretion results for different 

chronic intake periods are not independent for type S materials. For example, an intake in the 1950s 

could contribute to urinary excretion in the 1980s and later. To avoid potential underestimation of 

intakes for employees who worked at Hanford for relatively short periods, each chronic intake of type 

S material was independently fit using only the bioassay results from the single intake period. This 

same process was used for type M plutonium and 147Pm as well. This method likely results in an 

overestimate of intakes, particularly for assumed type S exposures that extend through multiple 

assumed intake periods.

C.4.3.1 Plutonium

Type M, S and SS materials were present at Hanford and were common enough to apply to general 

workers, therefore, all types were modeled. This section provides intakes for 239+240Pu of each type.

Dose reconstructions must also include intakes of other plutonium isotopes and 241Am as shown in 

Table 5-2. Plutonium-239+240 intakes are treated as 100% 239Pu for purposes of calculating the 

plutonium mixture.

C.4.3.1.1 Type M

This section shows intakes when the plutonium urinalysis results were fit as type M material. As 

discussed above, each chronic intake period was fit independently. Figures C-7 to C-10 show the fits 

to the 50th-percentile values for each intake period. The same intake periods were applied to the 

84th-percentile values because the values followed a similar pattern. Table C-13 summarizes the 

intake periods and corresponding intake rates for the 50th- and 84th-percentile values. Figure C-11 

depicts the predicted excretion rates from all 50th-percentile type M intakes.

Table C-13. Type M 239+240Pu intake periods and rates (dpm/d). 

Start Stop 50th-percentile 84th-percentile GSD

09/01/1946 12/31/1948 68.55 177.1 2.58

01/01/1949 12/31/1952 21.1 54.11 2.56

01/01/1953 12/31/1981 1.003 3.196 3.19

01/01/1982 12/31/1988 0.1336 1.132 8.47

The large decrease in 50th-percentile bioassay results and intakes after 1981 resulted from a switch 

in reporting practices from a reporting level of 0.025 dpm to reporting all results. The MDA also 

decreased from 0.05 dpm to 0.02 dpm in 1984.
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C.4.3.1.2 Type S

This section shows intakes when the plutonium urinalysis results were fit as type S material. As 

discussed above, each chronic intake period was fit independently. Figures C-12 to C-15 show the fits 

to the 50th-percentile values for each intake period. The same intake periods were applied to the 

84th-percentile values because the values followed a similar pattern. Table C-14 summarizes the 

intake period and corresponding intake rate for the 50th- and 84th-percentile values. Figure C-16 

depicts the predicted excretion rates from all 50th-percentile type S intakes.

Table C-14. Type S 239+240Pu intake period and rates (dpm/d). 

Start Stop 50th-percentile 84th-percentile GSD

09/01/1946 12/31/1948 2,122 5,504 2.59

01/01/1949 12/31/1952 495.8 1,270 2.56

01/01/1953 12/31/1981 12.21 38.82 3.18

01/01/1982 12/31/1988 2.654 20.45 7.71

The large decrease in 50th-percentile bioassay results and intakes after 1981 resulted from a switch 

in reporting practices from a reporting level of 0.025 dpm to reporting all results. The MDA also 

decreased from 0.05 dpm to 0.02 dpm in 1984.

C.4.3.1.3 Type SS 

This section shows intakes when the plutonium urinalysis results were fit as type SS material. As 

discussed above, each chronic intake period was fit independently. Figures C-17 to C-20 show the fits 

to the 50th-percentile values for each intake period. The same intake periods were applied to the 

84th-percentile values because the values followed a similar pattern. Table C-15 summarizes the 

intake period and corresponding intake rate for the 50th- and 84th-percentile values. Figure C-21 

depicts the predicted excretion rates from all 50th-percentile type SS intakes. 

Table C-15. Type SS 239+240Pu intake period and rates (dpm/d). 

Start Stop 50th-percentile 84th-percentile GSD

09/01/1946 12/31/1948 9,470 24,400 2.58

01/01/1949 12/31/1952 2,950 7,580 2.57

01/01/1953 12/31/1981 119 379 3.18

01/01/1982 12/31/1988 18.6 157 8.44

The large decrease in 50th-percentile bioassay results and intakes after 1981 resulted from a switch 

in reporting practices from a reporting level of 0.025 dpm to reporting all results. The MDA also 

decreased from 0.05 dpm to 0.02 dpm in 1984.

C.4.3.2 Uranium

Uranium at Hanford existed as mostly type F or S, but neither classification should be considered a 

perfect match to the uranium compounds, and type M cannot be ruled out. Therefore, all three types 

should be considered possible. Thorium exposure was associated with uranium exposure in 300 Area 

facilities from 1950 to 1970. Section C.5.2 contains guidance for adding thorium to uranium intakes. 

C.4.3.2.1 Type F

This section shows intakes when uranium urinalysis results were fit using a type F material.

Figure C-22 shows the fit to the 50th-percentile values from all intakes. The figure depicts the 
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expected excretion rate from an individual who was exposed for all the periods at the 50th-percentile 

intake rate. The same intake periods were applied to the 84th-percentile values because the values 

followed a similar pattern. Table C-16 summarizes the intake periods and corresponding intake rates 

for the 50th- and 84th-percentile values.

Table C-16. Type F uranium intake periods and rates (μg/d). 

Start Stop 50th-percentile 84th-percentile GSD

01/01/1948 12/31/1952 17.42 58.35 3.35

01/01/1953 12/31/1956 12.3 39.45 3.21

01/01/1957 12/31/1961 17.73 61.65 3.48

01/01/1962 12/31/1974 10.61 31.09 2.93

01/01/1975 12/31/1983 1.281 7.037 5.49

01/01/1984 12/31/1988 0.4903 1.895 3.86

The decrease in intake rate in 1975 resulted from a decrease in the reporting level from 4 μg/L to 

0.4 μg/L. That decrease could also have resulted in part from the shutdowns of the PUREX and UO3 

plants from 1973 to 1983.

C.4.3.2.2 Type M

The intake periods for the type F fits were applied to the type M material fits. Figure C-23 shows the fit 

to the 50th-percentile values from all intakes. The same intake periods were applied to the 

84th-percentile values because the values followed a similar pattern. Table C-17 summarizes the 

intake periods and corresponding intake rates for the 50th- and 84th-percentile values.

Table C-17. Type M uranium intake periods and rates (μg/d). 

Start Stop 50th-percentile 84th-percentile GSD

01/01/1948 12/31/1952 73.12 244.5 3.34

01/01/1953 12/31/1956 49.35 159 3.22

01/01/1957 12/31/1961 72.79 252.6 3.47

01/01/1962 12/31/1974 42.78 124.9 2.92

01/01/1975 12/31/1983 4.414 25.99 5.89

01/01/1984 12/31/1988 1.919 7.043 3.67

C.4.3.2.3 Type S

The intake periods for the type F and M fits were applied to the type S material fits. As discussed, 

each chronic intake period for type S material was fit independently. Figures C-24 to C-29 show the 

individual fits for the 50th-percentile values. The 84th-percentile values were fit similarly. Table C-18

summarizes the intake rates for the 50th- and 84th-percentile values. Figure C-30 shows the 

predicted excretion rates from all 50th-percentile value type S intakes.

Table C-18. Type S uranium intake periods and rates (μg/d). 

Start Stop 50th-percentile 84th-percentile GSD

01/01/1948 12/31/1952 1,308 4,373 3.34

01/01/1953 12/31/1956 1,035 3,321 3.21

01/01/1957 12/31/1961 1,349 4,660 3.45

01/01/1962 12/31/1974 665.1 1,953 2.94

01/01/1975 12/31/1983 83.67 403.3 4.82

01/01/1984 12/31/1988 46.84 171.6 3.66
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C.4.3.3 Strontium-90

Only type F solubility was analyzed for 90Sr. Figure C-31 shows the fit to the 50th-percentile values 

from all intakes. Table C-19 summarizes the intake rates for the 50th- and 84th-percentile values.

Table C-19. Type F 90Sr intake periods and rates (dpm/d). 

Start Stop 50th-percentile 84th-percentile GSD

01/01/1965 12/31/1967 90.47 241.5 2.67

01/01/1968 12/31/1970 60.5 241.5 3.99

01/01/1971 12/31/1988 3.698 9.405 2.54

C.4.3.4 Promethium-147

C.4.3.4.1 Type M

Each chronic intake period for type M 147Pm was fit independently. Figures C-32 and C-33 show the 

individual fits for the 50th-percentile values. The 84th-percentile values were fit similarly. Table C-20

summarizes the intake rates for the 50th- and 84th-percentile values. Figure C-34 shows the 

predicted excretion rates from all 50th-percentile value type M intakes.

Table C-20. Type M 147Pm intake periods and rates (dpm/d). 

Start Stop 50th-percentile 84th-percentile GSD

01/01/1966 12/31/1969 4,976 15,620 3.14

01/01/1970 12/31/1979 1,720 4,483 2.61

C.4.3.4.2 Type S

Each chronic intake period for type S 147Pm was fit independently. Figures C-35 and C-36 show the 

individual fits for the 50th-percentile values. The 84th–percentile values were fit similarly. Table C-21

summarizes the intake rates for the 50th- and 84th-percentile values. Figure C-37 shows the 

predicted excretion rates from all 50th-percentile value type S intakes.

Table C-21. Type S 147Pm intake periods and rates (dpm/d). 

Start Stop 50th-percentile 84th-percentile GSD

01/01/1966 12/31/1969 131,500 411,400 3.13

01/01/1970 12/31/1979 34,810 90,850 2.61

C.4.3.5 Zinc-65

Intakes of 65Zn include a single acute intake in addition to the normal chronic intakes that were 

modeled. This acute intake was included to optimize the fit of the modeling results to the data. The 

data show a sudden increase in the last quarter of 1976 that is followed by a steady decrease over 

the next several quarters, which is indicative of an acute intake during the last quarter of 1976.

C.4.3.5.1 Inhalation Intakes

Absorption type S was used for 65Zn inhalation intakes. Each chronic intake period for type S 65Zn was 

fit independently. However, the acute intake on October 1, 1976, was modeled with the 

contemporaneous chronic intake. Figures C-38 to C-43 show the individual fits for the 50th-percentile 

values. The 84th-percentile values were fit similarly. Table C-22 summarizes the intake rates for the 
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50th- and 84th-percentile values. Figure C-44 shows the predicted excretion rates from all 

50th-percentile value type S intakes.

Table C-22. Type S 65Zn inhalation intake periods and rates (pCi/d). 

Start Stop 50th-percentile 84th-percentile GSD

01/01/1960 12/31/1963 44.56 181.2 4.07

01/01/1964 12/31/1967 53.68 199.6 3.72

01/01/1968 12/31/1970 35.29 124.2 3.52

01/01/1971 12/31/1972 29.88 94.31 3.16

01/01/1973 12/31/1977 12.58 42.59 3.39

01/01/1978 12/31/1984 7.411 21.83 2.95

10/01/1976a 10/01/1976a 2,555b 14,260b 5.58

a. This is the date of the acute intake.

b. This intake has units of pCi for an acute intake rather than pCi/d.

C.4.3.5.2 Ingestion Intakes

The ingestion intakes of 65Zn were modeled together. Figure C-45 shows the fit to the 50th-percentile 

values from all intakes. The 84th-percentile values were fit similarly. Table C-23 summarizes the 

intake rates for the 50th- and 84th-percentile values. The once-through-cooled reactors were being 

shut down from 1964 to 1971 [ORAUT 2010a]. Ingestion intakes after the once-through-cooled 

reactors were shut down (after 1972) were not considered plausible.

Table C-23. Zinc-65 ingestion intake periods and rates (pCi/d). 

Start Stop 50th-percentile 84th-percentile GSD

01/01/1960 12/31/1963 26.21 105.0 4.01

01/01/1964 12/31/1967 27.64 103.4 3.74

01/01/1968 12/31/1970 16.69 57.34 3.44

01/01/1971 12/31/1972 13.06 39.86 3.05

C.4.3.6 Sodium-24

C.4.3.6.1 Inhalation Intakes

Absorption type F was used for 24Na inhalation intakes. Figure C-46 shows the fit to the 

50th-percentile values from all intakes. The 84th-percentile values were fit similarly. Table C-24 

summarizes the intake rates for the 50th- and 84th-percentile values.

Table C-24. Type F 24Na inhalation intake periods and rates 

(pCi/d). 

Start Stop 50th-percentile 84th-percentile GSD

01/01/1960 09/30/1963 310.3 942.6 3.04

10/01/1963 12/31/1963 1921 6,939 3.61

01/01/1964 09/30/1964 452.7 1,410 3.11

10/01/1964 03/31/1965 1,037.6 3,654 3.52

04/01/1965 12/31/1973 452.7 1,410 3.11

01/01/1974 12/31/1984 250.8 680.7 2.71
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C.4.3.6.2 Ingestion Intakes

The ingestion intakes of 24Na were modeled together. Figure C-47 shows the fit to the 50th-percentile 

values from all intakes. These depict the expected excretion rates from an individual who was

exposed for all the periods at the 50th- and 84th-percentile intake rates.

Table C-25 summarizes the intake rates for the 50th- and 84th-percentile values. Ingestion intakes 

after the once-through-cooled reactors were shut down (by 1971 because of the very short half-life of 

24Na) were not considered plausible.

Table C-25. Sodium-24 ingestion intake periods and rates (pCi/d). 

Start Stop 50th-percentile 84th-percentile GSD

01/01/1960 09/30/1963 205.6 624.7 3.04

10/01/1963 12/31/1963 1,274 4,599 3.61

01/01/1964 09/30/1964 311.9 986.3 3.16

10/01/1964 03/31/1965 687.6 2,421 3.52

04/01/1965 12/31/1971 311.9 986.3 3.16

C.4.3.7 Cesium-137

Only absorption type F was analyzed for 137Cs. Figure C-48 shows the fit to the 50th-percentile values 

from all intakes. Table C-26 summarizes the intake rates for the 50th- and 84th-percentile values.

Table C-26. Type F 137Cs intake periods and rates (pCi/d). 

Start Stop 50th-percentile 84th-percentile GSD

01/01/1960 12/31/1960 118.6 173.7 1.46

01/01/1961 12/31/1961 39.08 61.74 1.58

01/01/1962 12/31/1962 55.23 81.79 1.48

01/01/1963 12/31/1964 246.8 374.8 1.52

01/01/1965 12/31/1966 132.6 213.9 1.61

01/01/1967 12/31/1967 44.34 85.13 1.92

01/01/1968 12/31/1976 28.38 58.36 2.06

01/01/1977 12/31/1988 10.13 26.08 2.57

C.5 ASSIGNMENT OF INTAKES AND DOSES

Tables C-27 to C-36 summarize the 50th-percentile intake rates, GSDs, and 95th percentile intake 

rates for each primary radionuclide over their respective periods and solubility classes. When 

calculating doses to individuals from bioassay data, a GSD of 3 has been used to account for 

biological variation and uncertainty in the models. It was considered inappropriate to assign a value 

less than 3 for the co-exposure data. Therefore, a GSD of at least 3 was assigned for each of the 

intake periods for the 50th percentile intakes [ORAUT 2018]. The GSDs for different intake periods 

have also been adjusted in a manner favorable to claimants for consistency between intake periods 

for calculational efficiency. The 95th percentile intakes should be assigned as a constant rather than 

lognormal distribution.

The bioassay results that were used to establish co-exposure intakes did not cover all the years of 

possible exposure at Hanford. Extrapolation from monitored years to unmonitored years is discussed 

as part of the overall discussion of intakes for each radionuclide. The history of production at Hanford 

is relevant to those discussions. Figure C-6 shows how a figure of merit for production (tons of 

uranium processed in the reactors times the burnup) varied from 1944 to 1972. This unit is especially 
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Figure C-6. Production rates in tons of uranium times fuel burnup, 1944 to 1972 [136]. 

0

1000000

2000000

3000000

4000000

5000000

6000000

1940 1945 1950 1955 1960 1965 1970 1975

Year

Tons U times burnup

Production

relevant for the production of fission products and to a reasonable extent for production of activation 

products. The figure shows that the period of highest production at Hanford was from about 1958 to 

1968, which overlaps with the period of whole-body counting and 90Sr urinalysis.

For assignment of co-exposure intakes after 1988, the intake rates for 1988 for each radionuclide may 

be assigned. Given that the internal dosimetry and radiation safety programs generally improved with 

time, especially after 1988, the assignment of the 1988 intake rates for periods after 1988 is 

considered to be favorable to claimants. For radionuclides with no assigned intake rate in 1988, no 

intake post-1988 should be assigned.

C.5.1 Plutonium

Dose reconstructors should assume a 12% 239Pu (fuel-grade) mixture for co-exposure intakes in 

accordance with Table C-27. Tables C-27 through C-29 list the 239Pu intakes and associated GSDs for 

each period for type M, S, and SS solubility classes, respectively. A small amount of plutonium was 

produced in 1944 (T Plant started operations in December 1944), and more was produced in 1945. As 

presented in Section 5.1.3, an SEC was designated in which NIOSH found that it was infeasible to 

reconstruct doses for plutonium before August 31, 1946, due to inadequate monitoring. Therefore, 

dose reconstructions will not include any doses from plutonium before August 31, 1946.
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INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Table C-27. Type M 239Pu intake periods and rates (dpm/d). 

Start Stop 50th-percentile GSD 95th-percentile

09/01/1946 12/31/1948 68.55 3.00 418

01/01/1949 12/31/1952 21.1 3.00 129

01/01/1953 12/31/1981 1.003 3.19 6.76

01/01/1982 12/31/1988 0.1336 8.47 4.49

01/01/1989 Present 0.1336 8.47 4.49

Table C-28. Type S 239Pu intake period and rates (dpm/d). 

Start Stop 50th-percentile GSD 95th-percentile

09/01/1946 12/31/1948 2,122 3.00 12,930

01/01/1949 12/31/1952 495.8 3.00 3,021

01/01/1953 12/31/1981 12.21 3.18 81.9

01/01/1982 12/31/1988 2.654 7.71 76.4

01/01/1989 Present 2.654 7.71 76.4

Table C-29. Type SS 239Pu intake period and rates (dpm/d). 

Start Stop 50th-percentile GSD 95th-percentile

09/01/1946 12/31/1948 9,470 3.00 57,705

01/01/1949 12/31/1952 2,950 3.00 17,976

01/01/1953 12/31/1981 119 3.18 800.1

01/01/1982 12/31/1988 18.6 8.44 621.5

01/01/1989 Present 18.6 8.44 621.5

C.5.2 Uranium

Tables C-30 to C-32 list the uranium intakes and associated GSDs for each period. The values were 

adjusted from micrograms per day to picocuries per day for input into IMBA.

Table C-30. Uranium type F intake periods and rates (pCi/d). 

Start Stop

50th-percentile 

intake GSD

95th-percentile 

intake

12/01/1943 12/31/1948 Not applicable Not applicable Not applicable

01/01/1949 12/31/1951 12.29 3.48 95.6

01/01/1952 12/31/1952 15.85 3.48 123

01/01/1953 12/31/1956 11.19 3.48 87.0

01/01/1957 12/31/1961 16.13 3.48 125

01/01/1962 12/31/1974 9.654 3.00 58.8

01/01/1975 12/31/1983 1.166 5.49 19.2

01/01/1984 12/31/1988 0.446 3.86 4.11

01/01/1989 Present 0.446 3.86 4.11

Table C-31. Uranium type M intake periods and rates (pCi/d). 

Start Stop

50th-percentile 

Intake GSD

95th-percentile 

intake

12/01/1943 12/31/1948 See [BT 2011] 5 Not applicable

01/01/1949 12/31/1951 51.6 3.47 399

01/01/1952 12/31/1952 66.5 3.47 515

01/01/1953 12/31/1956 44.9 3.47 348

01/01/1957 12/31/1961 66.2 3.47 513

01/01/1962 12/31/1974 38.9 3.00 237

01/01/1975 12/31/1983 4.02 5.89 74.2
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Start Stop

50th-percentile 

Intake GSD

95th-percentile 

intake

01/01/1984 12/31/1988 1.75 3.67 14.8

01/01/1989 Present 1.75 3.67 14.8

Table C-32. Uranium type S intake periods and rates (pCi/d). 

Start Stop

50th-percentile 

Intake GSD

95th-percentile 

Intake

12/01/1943 12/31/1948 See [BT 2011] 5 Not applicable

01/01/1949 12/31/1951 923 3.45 7,078

01/01/1952 12/31/1952 1,190 3.45 9,126

01/01/1953 12/31/1956 942 3.45 7,224

01/01/1957 12/31/1961 1,227 3.45 9,409

01/01/1962 12/31/1974 605 3.00 3,686

01/01/1975 12/31/1983 76.1 4.82 1,012

01/01/1984 12/31/1988 42.6 3.66 361

01/01/1989 Present 42.6 3.66 361

Uranium bioassay started in 1947, but the early results were not considered reliable. The first 

extruded uranium rods arrived in October 1943 and machining began in December. From December 

1943 through 1948, intakes for uranium should be assigned in accordance with the Process/Job Title 

listed as Machining/Operator as presented in Battelle-TBD-6000 [BT 2011]. 

Starting in 1952, uranium at Hanford was RU with impurities. Impurities must be included with intakes 

of uranium; Section 5.2.5 contains the activities of impurities to apply to the uranium intakes in 

Tables C-28 to C-30 from 1952 to the present. The absorption types for the uranium impurities are 

provided in ORAUT-OTIB-0060, Internal Dose Reconstruction [ORAUT 2018]. 

C.5.3 Strontium-90

Table C-33 lists the 90Sr intakes and associated GSDs for each period. As presented in Section 5.1.3, 

an SEC was designated in which NIOSH found that it was infeasible to reconstruct doses for fission 

products before August 31, 1946, due to inadequate monitoring. Therefore, dose reconstructions will 

not include any doses from 90Sr before August 31, 1946. Production of fission products in general 

from 1965 to 1967 was as great as at any time in the history of Hanford, and buildup of 90Sr as a 

contamination source had occurred over 20 years of operation. Air concentration limits were 

essentially unchanged from 09/01/1946 to the present [Parker 1947, Patterson 1949]. Therefore, it is 

reasonable to apply the 90Sr intake rates from the 1965 to 1967 period to the 09/01/1946 to 1964 

period. 

Table C-33. Type F 90Sr intake periods and rates (dpm/d). 

Start Stop

50th-percentile 

intake GSD

95th-percentile 

Intake

09/01/1946 12/31/1964 90.47a 3.00 551.2

01/01/1965 12/31/1967 90.47 3.00 551.2

01/01/1968 12/31/1970 60.5 3.99 589

01/01/1971 12/31/1988 3.698 3.00 22.5

01/01/1989 Present 3.698 3.00 22.5

a. Intake rate assumed to be the same as the rate for 1965 to 1967.
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Review of claimant mixed fission product bioassay data for the 1958 through 1963 period revealed an 

essentially constant GM urinary excretion rate, which supports the assumption that use of the 1965 to 

1967 data is appropriate for 1947 through 1964.

C.5.4 Promethium-147

Tables C-34 and C-35 list respectively the Type M and Type F 147Pm intakes and associated GSDs 

for each period. Promethium-147 work occurred mostly in the 325 Building starting in 1961, 222-S 

from October 1964 through sometime in 1965, 201-C from 1964 through 1967, the 308 and probably 

the 309 Buildings approximately 1965 into the 1970s, and the 327 Building during major production 

(assume 1966–1970). See the Hanford Site Description for details [ORAUT 2010a]. Insufficient 

bioassay data were available for statistical analysis until 1966; intakes for unmonitored potentially 

exposed workers before 1966 were set equal to the 1966 to 1969 intakes because, for the most part, 

the same buildings and same facilities within those buildings (and in general even the same workers) 

were used for both periods. In addition, intakes from a February 1963 contamination spread were 

analyzed and found to be comparable at the 95th percentile to the 1966 to 1969 intake rates [137].

Table C-34. Type M 147Pm intake periods and rates (dpm/d). 

Start Stop

50th-percentile 

intake GSD

95th-percentile 

intake

10/01/1961 12/31/1965a 4,976 3.14 32,684

01/01/1966 12/31/1969 4,976 3.14 32,684

01/01/1970 12/31/1979 1,720 3.00 10,481

a. Intakes set equal to the 1966 to 1969 intakes.

Table C-35. Type S 147Pm intake periods and rates (dpm/d). 

Start Stop

50th-percentile 

intake GSD

95th-percentile 

intake

10/01/1961 12/31/1965a 131,500 3.13 859,204

01/01/1966 12/31/1969 131,500 3.13 859,204

01/01/1970 12/31/1979 34,810 3.00 212,114

a. Intakes set equal to the 1966 to 1969 intakes.

C.5.5 Zinc-65

Tables C-36 and C-37 list respectively the inhalation and ingestion zinc intakes and associated GSDs 

for each period. As presented in Section 5.1.3, an SEC was designated in which NIOSH found that it 

was infeasible to reconstruct doses for fission products before August 31, 1946, due to inadequate 

monitoring. Therefore, dose reconstructions will not include any doses from 65Zn before August 31, 

1946. Similar to 90Sr, the intake rates before 1960 were assumed based on smaller production and 

similar air concentration limits. Review of claimant mixed fission product bioassay data for the 1958 

through 1963 period revealed an essentially constant GM urinary excretion rate, which supports the 

assumption that use of the 1960 to 1963 data is appropriate for 09/01/1946 through 1959. Because 

the intakes are based on whole-body counting, which measures activity in the body from all intake 

pathways, the dose reconstructor should apply the more claimant-favorable of the inhalation or 

ingestion intakes, but not both in the same year. If ingestion is used and the exposure period

continues to 1973 or later, the intake mode must be switched from ingestion to inhalation for the latter 

years. Because ingestion was determined not to be plausible after 1972, the intakes for inhalation 

must be assumed after this date, even when ingestion has been assumed before this date.
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Table C-36. Zinc-65 inhalation intake periods and rates (pCi/d). 

Start Stop

50th-percentile 

intake GSD

95th-percentile 

intake

09/01/1946 12/31/1959a 44.56 4.07 448

01/01/1960 12/31/1963 44.56 4.07 448

01/01/1964 12/31/1967 53.68 4.07 540

01/01/1968 12/31/1970 35.29 3.52 280

01/01/1971 12/31/1972 29.88 3.52 237

01/01/1973 12/31/1977 12.58 3.52 99.7

01/01/1978 12/31/1984 7.411 3.00 45.2

10/01/1976b 10/01/1976 2,555c 5.58 43,212

a. Intake rates assumed to be the same as the rates for 1960 to 1963.

b. This is the date of the acute intake.

c. This intake has units of pCi for an acute intake rather than pCi/d.

Table C-37. Zinc-65 ingestion intake periods and rates (pCi/d). 

Start Stop

50th-percentile 

intake GSD

95th-percentile 

intake

09/01/1946 12/31/1959a 26.2 4.01 257

01/01/1960 12/31/1963 26.2 4.01 257

01/01/1964 12/31/1967 27.64 4.01 271

01/01/1968 12/31/1970 16.69 3.44 127

01/01/1971 12/31/1972 13.06 3.05 82

01/01/1973 12/31/1977 Not applicable Not applicable Not applicable

01/01/1978 12/31/1984 Not applicable Not applicable Not applicable

10/01/1976b 10/01/1976 Not applicable Not applicable Not applicable

a. Intake rates assumed to be the same as the rates for 1960 to 1963.

b. This is the date of the acute intake.

C.5.6 Sodium-24

Tables C-38 and C-39 list respectively the type F inhalation and ingestion 24Na intakes and associated 

GSDs for each period. As an assumption favorable to claimants, the highest intake rate within a given 

year was used to set the intake rate for the entire year. As presented in Section 5.1.3, an SEC was

designated in which NIOSH found that it was infeasible to reconstruct doses for fission products

before August 31, 1946, due to inadequate monitoring. Therefore, dose reconstructions will not 

include any doses from 24Na before August 31, 1946. Like 90Sr, the intake rates before 1960 were 

assumed based on smaller production and similar air concentration limits with the exception of the 

1944 to 08/31/1946 period, which was supported by the review of the fission product urine data as 

discussed previously. Because the intakes were based on whole-body counting, which measures 

activity in the body from all intake pathways, the dose reconstructor should apply the more claimantfavorable of the inhalation or ingestion intakes, but not both in the same year. 

If ingestion is used and the exposure period continues to 1972 or later, the intake mode must be 

switched from ingestion to inhalation for the latter years; that is, the intakes for inhalation must be 

assumed after 1971, even when ingestion has been assumed before this date.
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Table C-38. Type F 24Na inhalation intake periods and rates (pCi/d). 

Start Stop 50th-percentile intake GSD 95th-percentile intake

09/01/1946 12/31/1959a 310.3 3.04 1,932

01/01/1960 12/31/1962 310.3 3.04 1,932

01/01/1963 12/31/1963 1,921 3.61 15,872

01/01/1964 12/31/1965 1,038 3.52 8,228

01/01/1966 12/31/1971 452.7 3.11 2,927

01/01/1972 12/31/1973 452.7 3.11 2,927

01/01/1974 12/31/1984 250.8 3.11 1,622

a. Intake rates assumed to be the same as the rates for 1960 to 1962.

Table C-39. Type F 24Na ingestion intake periods and rates (pCi/d). 

Start Stop

50th-percentile 

intake GSD

95th-percentile 

intake

09/01/1946 12/31/1959a 205.6 3.04 1,280

01/01/1960 12/31/1962 205.6 3.04 1,280

01/01/1963 12/31/1963 1,274 3.61 10,526

01/01/1964 12/31/1965 687.6 3.61 5,681

01/01/1966 12/31/1971 311.9 3.61 2,577

01/01/1972 12/31/1973 Not applicable Not applicable Not applicable

01/01/1974 12/31/1984 Not applicable Not applicable Not applicable

a. Intake rates assumed to be the same as the rates for 1960 to 1962.

C.5.7 Cesium-137

Table C-40 lists the 137Cs intakes and associated GSDs for each period. As presented in Section 

5.1.3, an SEC was designated in which NIOSH found that it was infeasible to reconstruct doses for 

fission products before August 31, 1946, due to inadequate monitoring. Therefore, dose 

reconstructions will not include any doses from 137Cs before August 31, 1946. Similar to 90Sr, the 

intake rates before 1960 were assumed based on smaller production and similar air concentration 

limits, with the exception of the 1944 to 08/31/1946 period. 

Table C-40. Type F 137Cs intake periods and rates (pCi/d). 

Start Stop

50th-percentile 

intake GSD

95th-percentile 

intake

09/01/1946 12/31/1959 118.6a 3.00 723

01/01/1960 12/31/1960 118.6 3.00 723

01/01/1961 12/31/1961 39.08 3.00 238

01/01/1962 12/31/1962 55.23 3.00 337

01/01/1963 12/31/1964 246.8 3.00 1,503

01/01/1965 12/31/1966 132.6 3.00 808

01/01/1967 12/31/1967 44.34 3.00 270

01/01/1968 12/31/1976 28.38 3.00 173

01/01/1977 12/31/1988 10.13 3.00 61.7

01/01/1989 Present 10.13 3.00 61.7

a. Intake rate assumed to be the same as the rate for 1960.

C.5.8 Additional Radionuclides

A large number of different radionuclides were present at Hanford at various times, but the available

bioassay data for radionuclides in addition to those in this attachment were considered to be too few 

to be statistically reliable for intake estimation. Workers who were exposed to 137Cs, 90Sr, 24Na, and 
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65Zn could also have been exposed to other fission and activation products. From 1960 to 1988, 

intakes of most fission or activation products would have been detectable in whole-body counts, but 

the recording practice for fission and activation products other than 137Cs, 24Na, and 65Zn was not 

amenable to statistical analysis. Guidance for assigning intakes of additional radionuclides is provided 

in ORAUT-OTIB-0054 [ORAUT 2015] and for specific buildings in Section 5.6. 

Assignment to unmonitored workers of radionuclides that were not analyzed in this study should be in 

accordance with guidance in Section 5.6. 

C.6 STATISTICAL SUMMARIES AND PLOTS

In the Tables C-41 to C-47, the GM is the 50th-percentile value of the fitted line and GM × GSD is the 

84th-percentile value.

In Figures C-7 to C-48, the annual bioassay data that were used in the fits are shown as solid blue 

dots (●) (dark dots when printed in black and white), and data that were not used in the fits are shown 

as red dots (●) (light dots when printed in black and white). Tables C-48 and C-49 list the references 

for the data.

Table C-41. Statistical summary of 239Pu 24-hour urinary 

excretion rates (dpm/d), 1946 to 1988.

a

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

11/01/1946 0.25119 0.64203

07/01/1947 0.20476 0.50991

04/01/1948 0.21047 0.53721

10/01/1948 0.22645 0.60437

07/01/1949 0.10752 0.28933

07/01/1950 0.10249 0.25412

07/01/1951 0.10358 0.26263

07/01/1952 0.05707 0.14968

04/01/1953 0.01647 0.04436

10/01/1953 0.01724 0.04873

02/15/1954 0.01704 0.04637

05/15/1954 0.01746 0.04932

08/15/1954 0.01765 0.05080

11/15/1954 0.01697 0.04609

02/15/1955 0.01671 0.04478

05/15/1955 0.01695 0.04624

08/15/1955 0.01746 0.05021

11/15/1955 0.01681 0.04538

02/15/1956 0.01703 0.04723

05/15/1956 0.01716 0.04803

08/15/1956 0.01724 0.04828

11/15/1956 0.01756 0.04956

02/15/1957 0.01659 0.04369

05/15/1957 0.01699 0.04599

08/15/1957 0.01663 0.04360

11/15/1957 0.01712 0.04671

02/15/1958 0.01576 0.06661

05/15/1958 0.01089 0.03745

08/15/1958 0.01001 0.03106
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INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

11/15/1958 0.01240 0.05037

02/15/1959 0.01265 0.05288

05/15/1959 0.01163 0.04018

08/15/1959 0.01058 0.03276

11/15/1959 0.01098 0.03388

02/15/1960 0.01237 0.04317

05/15/1960 0.01002 0.02983

08/15/1960 0.00980 0.02880

11/15/1960 0.00980 0.02931

02/15/1961 0.00977 0.02866

05/15/1961 0.00984 0.02971

08/15/1961 0.01022 0.03099

11/15/1961 0.00990 0.02813

02/15/1962 0.00948 0.02621

05/15/1962 0.01091 0.03307

08/15/1962 0.01235 0.04221

11/15/1962 0.01061 0.03407

02/15/1963 0.01100 0.03651

05/15/1963 0.01276 0.04553

08/15/1963 0.01166 0.03917

11/15/1963 0.01479 0.05371

02/15/1964 0.01826 0.06779

05/15/1964 0.02133 0.08054

08/15/1964 0.02435 0.08191

11/15/1964 0.02067 0.07232

02/15/1965 0.01409 0.04706

05/15/1965 0.01380 0.04542

08/15/1965 0.01248 0.03685

11/15/1965 0.01556 0.04811

02/15/1966 0.01197 0.03675

05/15/1966 0.01152 0.03613

08/15/1966 0.01354 0.04801

11/15/1966 0.01200 0.04220

02/15/1967 0.01330 0.05292

05/15/1967 0.02204 0.11370

08/15/1967 0.01175 0.04248

11/15/1967 0.01338 0.05001

02/15/1968 0.01216 0.04585

05/15/1968 0.01078 0.03891

08/15/1968 0.01397 0.06615

11/15/1968 0.01192 0.04489

02/15/1969 0.01321 0.04680

05/15/1969 0.00952 0.02753

08/15/1969 0.01101 0.03788

11/15/1969 0.01095 0.03868

02/15/1970 0.00974 0.03019

05/15/1970 0.00888 0.02567

08/15/1970 0.00980 0.03067

11/15/1970 0.01307 0.05338

02/15/1971 0.00933 0.02895

05/15/1971 0.00993 0.03206
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Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

08/15/1971 0.01126 0.03955

11/15/1971 0.01089 0.03976

02/15/1972 0.01179 0.04290

05/15/1972 0.01013 0.03254

08/15/1972 0.00973 0.03035

11/15/1972 0.01073 0.03642

02/15/1973 0.01107 0.03977

05/15/1973 0.00948 0.02877

08/15/1973 0.00945 0.02812

11/15/1973 0.00875 0.02403

02/15/1974 0.00885 0.02454

05/15/1974 0.00900 0.02532

08/15/1974 0.00911 0.02602

11/15/1974 0.01119 0.03852

02/15/1975 0.00910 0.02516

05/15/1975 0.00953 0.02738

08/15/1975 0.00802 0.02030

11/15/1975 0.00838 0.02233

02/15/1976 0.00857 0.02433

05/15/1976 0.00835 0.02245

08/15/1976 0.00801 0.02006

11/15/1976 0.00898 0.02764

02/15/1977 0.00871 0.02336

05/15/1977 0.00875 0.02371

08/15/1977 0.00835 0.02161

11/15/1977 0.00812 0.02061

02/15/1978 0.00878 0.02468

05/15/1978 0.00906 0.02664

08/15/1978 0.00836 0.02192

11/15/1978 0.00825 0.02180

02/15/1979 0.00871 0.02373

05/15/1979 0.00849 0.02278

08/15/1979 0.00851 0.02273

11/15/1979 0.00818 0.02095

02/15/1980 0.00821 0.02140

05/15/1980 0.00865 0.02388

08/15/1980 0.00870 0.02426

11/15/1980 0.00900 0.02488

02/15/1981 0.00799 0.02036

05/15/1981 0.00827 0.02204

08/05/1981 0.00851 0.02341

11/05/1981 0.00109 0.00644

02/15/1982 0.00069 0.00973

05/15/1982 0.00053 0.00572

08/15/1982 0.00083 0.00647

11/15/1982 0.00149 0.01120

02/15/1983 0.00113 0.00661

05/15/1983 0.00087 0.00497

08/15/1983 0.00094 0.00738

11/15/1983 0.00083 0.00589

02/15/1984 0.00132 0.00760
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Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

05/15/1984 0.00062 0.01534

08/15/1984 0.00090 0.00810

11/15/1984 0.00099 0.00860

02/15/1985 0.00116 0.01028

05/15/1985 0.00056 0.00531

08/15/1985 0.00076 0.00541

11/15/1985 0.00073 0.00389

02/15/1986 0.00065 0.00635

05/15/1986 0.00069 0.01717

08/15/1986 0.00099 0.00725

11/15/1986 0.00050 0.00463

02/15/1987 0.00069 0.00551

05/15/1987 0.00065 0.00471

08/15/1987 0.00086 0.00488

11/15/1987 0.00048 0.00411

02/15/1988 0.00080 0.00395

05/15/1988 0.00052 0.00312

08/15/1988 0.00064 0.00339

11/15/1988 0.00020 0.00203

a. For 1946 through 1949, the Pu-239 values were calculated as 0.824 

of the total Pu alpha values. For 1950 through 1954, the 

Pu-239 values were calculated as 0.830 of the total Pu alpha 

values. For 1951 through 08/15/1983, the Pu-239 values were 

calculated as 0.835 of the total Pu alpha values.

Figure C-7. 50th-percentile plutonium urinalysis data for intakes of 

type M material, 1946 to 1948. 

● = used in fit, ● = not used in fit, — = fit
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Figure C-8. 50th-percentile plutonium urinalysis data for intakes of 

type M material, 1949 to 1952. 

● = used in fit, ● = not used in fit, — = fit

Figure C-9. 50th-percentile plutonium urinalysis data for intakes of 

type M material, 1953 to 1981. 

● = used in fit, ● = not used in fit, — = fit

Figure C-10. 50th-percentile plutonium urinalysis data for intakes of 

type M material, 1982 to 1988. 

● = used in fit, ● = not used in fit, — = fit
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure C-11. Predicted 50th-percentile urinary excretion of type M 

plutonium from 1946 to 1988 based on four independent intakes, 

compared to bioassay data. 

● = used in fit, ● = not used in fit, — = fit

Figure C-12. 50th-percentile plutonium urinalysis data for intakes of 

type S material, 1946 to 1948. 

● = used in fit, ● = not used in fit, — = fit

Figure C-13. 50th-percentile plutonium urinalysis data for intakes of 

type S material, 1949 to 1952. 

● = used in fit, ● = not used in fit, — = fit
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure C-14. 50th-percentile plutonium urinalysis data for intakes of 

type S material, 1953 to 1981. 

● = used in fit, ● = not used in fit, — = fit

Figure C-15. 50th-percentile plutonium urinalysis data for intakes of 

type S material, 1982 to 1988. 

● = used in fit, ● = not used in fit, — = fit

Figure C-16. Predicted 50th-percentile urinary excretion of type S 

plutonium from 1946 to 1988 based on four independent intakes, 

compared to bioassay data. 

● = used in fit, ● = not used in fit, — = fit
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure C-17. 50th-percentile plutonium urinalysis data for intakes of type SS material, 1946 to 1948. 

● = used in fit, ● = not used in fit, — = fit

Figure C-18. 50th-percentile plutonium urinalysis data for intakes of type SS material, 1949 to 1952. 

● = used in fit, ● = not used in fit, — = fit
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure C-19. 50th-percentile plutonium urinalysis data for intakes of type SS material, 1953 to 1981. 

● = used in fit, ● = not used in fit, — = fit

Figure C-20. 50th-percentile plutonium urinalysis data for intakes of type SS material, 1982 to 1988. 

● = used in fit, ● = not used in fit, — = fit
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure C-21. Predicted 50th-percentile urinary excretion of type SS plutonium from 1946 to 1988 

based on four independent intakes, compared to bioassay data. 

● = used in fit, ● = not used in fit, — = fit

Table C-42. Statistical summary of uranium 24-hour

urinary excretion rates (μg/d), 1948 to 1988.

a 

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

04/01/1948 3.88 12.48

10/01/1948 5.50 21.41

02/15/1949 4.20 15.22

05/15/1949 3.32 10.87

08/15/1949 7.02 20.71

11/15/1949 4.65 15.47

02/15/1950 6.97 23.35

05/15/1950 4.57 15.20

08/15/1950 4.71 13.26

11/15/1950 3.62 11.81

02/15/1951 3.97 13.61

05/15/1951 4.15 14.93

08/15/1951 4.51 16.53

11/15/1951 4.15 13.99

02/15/1952 5.71 20.24

05/15/1952 4.84 16.17

08/15/1952 4.48 14.71

11/15/1952 5.15 16.08

02/15/1953 4.89 15.49

05/15/1953 4.07 12.44

08/15/1953 3.39 10.92

11/15/1953 3.62 12.79

02/15/1954 3.02 9.77

05/15/1954 3.02 9.26

08/15/1954 3.16 10.11

11/15/1954 3.78 12.74

02/14/1955 3.17 9.87

05/15/1955 3.44 11.06

08/15/1955 2.77 8.14
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

11/15/1955 3.21 10.35

02/14/1956 2.61 7.45

05/15/1956 3.55 11.48

08/15/1956 4.20 14.94

11/15/1956 2.88 8.92

02/15/1957 4.28 15.51

05/15/1957 5.70 21.26

08/15/1957 6.23 23.63

11/51/1957 6.88 24.31

02/15/1958 6.65 22.18

05/15/1958 5.34 18.07

08/15/1958 4.34 14.70

11/15/1958 3.96 12.54

02/15/1959 3.94 12.97

05/15/1959 4.59 15.29

08/15/1959 5.54 19.18

11/15/1959 4.48 16.23

02/15/1960 4.20 14.36

05/15/1960 5.82 20.97

08/15/1960 4.47 14.79

11/15/1960 4.34 15.45

02/15/1961 4.76 18.96

05/15/1961 4.85 17.54

08/15/1961 3.76 11.26

11/15/1961 3.86 11.32

02/15/1962 3.68 11.21

05/15/1962 3.05 8.70

08/15/1962 2.78 7.73

11/15/1962 2.74 7.69

02/15/1963 3.77 10.83

05/15/1963 3.19 8.55

08/15/1963 2.73 7.43

11/15/1963 2.76 7.37

02/15/1964 2.39 6.25

05/15/1964 2.54 6.79

08/15/1964 2.89 9.30

11/15/1964 2.66 7.54

02/15/1965 3.56 11.55

05/15/1965 3.82 13.26

08/15/1965 4.39 16.82

11/15/1965 3.72 11.94

02/15/1966 3.09 10.22

05/15/1966 3.02 9.28

08/15/1966 3.56 12.10

11/15/1966 3.13 9.35

02/15/1967 2.85 8.54

05/15/1967 2.97 8.84

08/15/1967 2.71 7.71

11/15/1967 2.77 8.13

02/15/1968 3.06 9.14

05/15/1968 3.73 11.62
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

08/15/1968 2.96 8.82

11/15/1968 3.21 9.10

02/15/1969 2.38 5.98

05/15/1969 2.73 7.06

08/15/1969 2.48 6.26

11/15/1969 2.27 5.67

07/01/1970 2.24 5.65

07/01/1971 2.61 6.88

07/01/1972 3.10 8.19

07/01/1973 2.27 5.34

07/01/1974 2.86 8.96

07/01/1975 0.97 5.99

07/01/1976 0.79 4.84

07/01/1977 0.39 1.42

07/01/1978 0.25 0.76

07/01/1979 0.29 0.96

07/01/1980 0.25 0.74

07/01/1981 0.41 2.02

07/01/1982 0.36 2.11

07/01/1983 0.13 0.64

02/15/1984 0.14 0.82

05/15/1984 0.20 0.78

08/15/1984 0.25 1.15

11/15/1984 0.32 1.34

02/15/1985 0.31 1.25

05/15/1985 0.28 0.89

08/15/1985 0.23 0.71

11/15/1985 0.15 0.45

02/15/1986 0.13 0.38

05/15/1986 0.27 1.36

08/15/1986 0.21 0.78

11/15/1986 0.18 0.62

02/15/1987 0.12 0.43

05/15/1987 0.15 0.50

08/15/1987 0.14 0.42

11/15/1987 0.15 0.43

02/15/1988 0.14 0.41

05/15/1988 0.09 0.23

08/15/1988 0.09 0.29

11/15/1988 0.02 0.30

a. Through 1981, excretion was converted to μg/d from μg/L assuming 

1.4L/d excretion.
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure C-22. 50th-percentile uranium urinalysis data for intakes of type 

F material, 1948 to 1988. 

● = used in fit, ● = not used in fit, — = fit

Figure C-23. 50th-percentile uranium urinalysis data for intakes of 

type M material, 1948 to 1988. 

● = used in fit, ● = not used in fit, — = fit

Figure C-24. 50th-percentile uranium urinalysis data for intakes of 

type S material, 1948 to 1952. 

● = used in fit, ● = not used in fit, — = fit
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure C-25. 50th-percentile uranium urinalysis data for intakes of 

type S material, 1953 to 1956. 

● = used in fit, ● = not used in fit, — = fit

Figure C-26. 50th-percentile uranium urinalysis data for intakes of 

type S material, 1957 to 1961. 

● = used in fit, ● = not used in fit, — = fit

Figure C-27. 50th-percentile uranium urinalysis data for intakes of 

type S material, 1962 to 1974. 

● = used in fit, ● = not used in fit, — = fit
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure C-28. 50th-percentile uranium urinalysis data for intakes of 

type S material, 1975 to 1983. 

● = used in fit, ● = not used in fit, — = fit

Figure C-29. 50th-percentile uranium urinalysis data for intakes of 

type S material, 1984 to 1988. 

● = used in fit, ● = not used in fit, — = fit

Figure C-30. Predicted 50th-percentile urinary excretion of type S 

uranium from 1946 to 1988 based on six independent intakes, 

compared to bioassay data. 

● = used in fit, ● = not used in fit, — = fit
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Table C-43. Statistical summary of 90Sr 24-hour urinary 

excretion rates (dpm/d), 1966 to 1988. 

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

07/01/1966 20.74 56.46

07/01/1969 14.53 55.88

07/01/1971 1.76 4.56

07/01/1972 2.33 12.27

07/01/1973 1.64 4.51

07/01/1974 1.38 3.82

07/01/1975 0.923 2.57

07/01/1976 1.20 3.40

07/01/1977 0.835 2.12

07/01/1978 1.009 3.82

02/15/1979 0.882 2.55

08/15/1979 2.00 5.58

07/01/1980 1.92 4.92

07/01/1981 2.33 7.60

02/15/1982 0.289 1.43

05/15/1982 0.322 1.58

08/15/1982 0.350 1.72

11/15/1982 0.802 7.91

02/15/1983 0.665 2.60

05/15/1983 0.387 3.13

08/15/1983 0.743 3.78

11/15/1983 0.410 2.73

02/15/1984 0.807 1.76

05/15/1984 1.04 2.70

08/15/1984 1.00 1.66

11/15/1984 1.06 1.87

02/15/1985 1.30 2.88

05/15/1985 1.18 2.38

08/15/1985 1.06 1.69

11/15/1985 1.11 2.20

02/15/1986 1.12 2.26

05/15/1986 1.05 1.88

08/15/1986 1.14 2.05

11/15/1986 1.27 2.29

02/15/1987 1.28 2.02

05/15/1987 1.43 2.28

08/15/1987 1.43 2.18

11/15/1987 1.47 2.21

02/15/1988 1.79 2.79

05/15/1988 1.53 2.57

08/15/1988 1.32 2.02

11/15/1988 0.327 0.993
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure C-31. 50th-percentile strontium urinalysis data for intakes of 

type F material, 1965 to 1988. 

● = used in fit, ● = not used in fit, — = fit

Table C-44. Statistical summary of 147Pm 24-hour 

urinary excretion rates (dpm/d), 1967 to 1977. 

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

01/01/1967 25.7 82.9

02/12/1969 28.6 88.5

02/15/1971 13.7 35.6

07/01/1977a 10.5 27.4

a. Very few bioassays were obtained during 1972 through 1974 

so no statistics were developed. A single intake rate was fit for 

1970 through 1979, using the 1971 and 1977 statistical 

values.

Figure C-32. 50th-percentile promethium urinalysis data for intakes of 

type M material, 1966 to 1969. 

● = used in fit, ● = not used in fit, — = fit
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure C-33. 50th-percentile promethium urinalysis data for intakes of 

type M material, 1970 to 1979. 

● = used in fit, ● = not used in fit, — = fit

Figure C-34. Predicted 50th-percentile urinary excretion of type M 

promethium from 1966 to 1979 based on two independent intakes, 

compared to bioassay data. 

● = used in fit, ● = not used in fit, — = fit

Figure C-35. 50th-percentile promethium urinalysis data for intakes of 

type S material, 1966 to 1969. 

● = used in fit, ● = not used in fit, — = fit
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure C-36. 50th-percentile promethium urinalysis data for intakes of 

type S material, 1970 to 1979. 

● = used in fit, ● = not used in fit, — = fit

Figure C-37. Predicted 50th-percentile urinary excretion of type S 

promethium from 1966 to 1979 based on two independent intakes, 

compared to bioassay data. 

● = used in fit, ● = not used in fit, — = fit

Table C-45. Statistical summary of 65Zn measured in 

whole-body counts (nCi), 1960 to 1984. 

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

07/01/1960 2.45 7.80

04/01/1961 2.16 9.58

10/01/1961 2.13 7.78

04/01/1962 2.04 8.73

10/01/1962 2.15 12.08

02/15/1963 2.11 8.44

05/15/1963 1.60 6.25

08/15/1963 1.34 4.70

11/15/1963 3.40 12.52

02/15/1964 2.40 9.02

05/15/1964 2.12 7.22

08/15/1964 1.92 6.79
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

11/15/1964 4.20 16.06

02/15/1965 2.93 12.17

05/15/1965 3.11 10.80

08/15/1965 2.348 9.40

11/15/1965 2.56 10.10

02/15/1966 2.82 9.32

05/15/1966 2.87 10.10

08/15/1966 1.84 6.32

11/15/1966 1.86 8.03

02/15/1967 1.78 6.06

05/15/1967 3.35 12.02

08/15/1967 2.13 8.26

11/15/1967 1.50 5.91

02/15/1968 1.53 6.61

05/15/1968 1.38 5.45

08/15/1968 1.41 6.18

11/15/1968 1.83 6.78

02/15/1969 2.35 7.89

05/15/1969 1.28 4.63

08/15/1969 1.05 3.66

11/15/1969 1.45 5.37

02/15/1970 1.45 5.56

05/15/1970 1.27 4.31

08/15/1970 1.56 5.11

11/15/1970 2.10 5.87

02/15/1971 1.52 5.06

05/15/1971 1.55 4.69

08/15/1971 1.29 3.81

11/15/1971 1.24 3.55

02/15/1972 1.18 3.92

05/15/1972 0.963 3.01

08/15/1972 1.17 3.74

11/15/1972 1.18 3.98

02/15/1973 1.18 4.21

05/15/1973 1.04 3.44

08/15/1973 0.612 1.99

11/15/1973 0.678 2.27

02/15/1974 0.497 1.61

05/15/1974 0.431 1.33

08/15/1974 0.456 1.41

11/15/1974 0.507 1.64

04/01/1975 0.405 1.23

10/01/1975 0.613 2.00

02/15/1976 0.647 2.19

05/15/1976 0.712 3.25

08/15/1976 0.673 2.46

11/15/1976 1.21 5.39

02/15/1977 0.964 3.74

05/15/1977 0.788 2.83

08/15/1977 0.688 2.67

11/15/1977 0.388 1.27
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

02/15/1978 0.371 1.16

05/15/1978 0.369 1.15

08/15/1978 0.373 1.19

11/15/1978 0.396 1.28

02/15/1979 0.401 1.28

05/15/1979 0.415 1.37

08/15/1979 0.348 1.02

11/15/1979 0.406 1.31

02/15/1980 0.384 1.19

05/15/1980 0.403 1.32

08/15/1980 0.372 1.15

11/15/1980 0.348 1.01

02/15/1981 0.372 1.09

05/15/1981 0.354 1.04

08/15/1981 0.306 0.817

11/15/1981 0.316 0.854

02/15/1982 0.340 0.947

05/15/1982 0.338 0.944

08/15/1982 0.334 0.934

11/15/1982 0.331 0.909

02/15/1983 0.323 0.884

05/15/1983 0.334 0.922

08/15/1983 0.314 0.843

11/15/1983 0.337 0.937

04/01/1984 0.354 1.03

10/01/1984 0.391 1.09

Figure C-38. 50th-percentile zinc whole-body counting data for 

inhalation intakes of type S material, 1960 to 1963. 

● = used in fit, ● = not used in fit, — = fit
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure C-39. 50th-percentile zinc whole-body counting data for 

inhalation intakes of type S material, 1964 to 1967. 

● = used in fit, ● = not used in fit, — = fit

Figure C-40. 50th-percentile zinc whole-body counting data for 

inhalation intakes of type S material, 1968 to 1970. 

● = used in fit, ● = not used in fit, — = fit

Figure C-41. 50th-percentile zinc whole-body counting data for 

inhalation intakes of type S material, 1971 to 1972. 

● = used in fit, ● = not used in fit, — = fit
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure C-42. 50th-percentile zinc whole-body counting data for 

inhalation intakes of type S material, 1973 to 1977. 

● = used in fit, ● = not used in fit, — = fit 

Figure C-43. 50th-percentile zinc whole-body counting data for 

inhalation intakes of type S material, 1978 to 1984. 

● = used in fit, ● = not used in fit, — = fit

Figure C-44. Predicted 50th-percentile urinary excretion of inhaled 

type S zinc from 1960 to 1984 based on six independent chronic 

intakes and one acute intake, compared to bioassay data. 

● = used in fit, ● = not used in fit, — = fit
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Figure C-45. 50th-percentile zinc whole-body counting data for 

ingestion intakes, 1960 to 1984. 

● = used in fit, ● = not used in fit, — = fit

Table C-46. Statistical summary of 24Na measured in 

whole-body counts (nCi), 1960 to 1984. 

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

07/01/1960 0.221 0.587

04/01/1961 0.174 0.610

10/01/1961 0.187 0.616

02/15/1962 0.142 0.421

05/15/1962 0.182 0.609

08/15/1962 0.150 0.413

11/15/1962 0.145 0.414

02/15/1963 0.180 0.562

05/15/1963 0.185 0.515

08/15/1963 0.231 0.713

11/15/1963 1.11 4.02

02/15/1964 0.388 1.35

05/15/1964 0.355 1.16

08/15/1964 0.316 0.993

11/15/1964 0.647 2.06

02/15/1965 0.555 2.17

05/15/1965 0.295 0.983

08/15/1965 0.286 0.930

11/15/1965 0.292 0.891

02/15/1966 0.347 1.09

05/15/1966 0.233 0.650

08/15/1966 0.257 0.834

11/15/1966 0.218 0.687

02/15/1967 0.288 0.904

05/15/1967 0.428 1.740

08/15/1967 0.247 0.701

11/15/1967 0.346 0.998

02/15/1968 0.405 1.610

05/15/1968 0.269 0.887

08/15/1968 0.229 0.728

11/15/1968 0.256 0.805
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

02/15/1969 0.194 0.570

05/15/1969 0.204 0.649

08/15/1969 0.268 0.915

11/15/1969 0.285 0.979

02/15/1970 0.218 0.623

05/15/1970 0.228 0.605

08/15/1970 0.193 0.543

11/15/1970 0.196 0.559

02/15/1971 0.260 0.711

05/15/1971 0.234 0.591

08/15/1971 0.224 0.589

11/15/1971 0.220 0.591

02/15/1972 0.245 0.715

05/15/1972 0.255 0.760

08/15/1972 0.206 0.608

11/15/1972 0.243 0.673

02/15/1973 0.244 0.756

05/15/1973 0.196 0.556

08/15/1973 0.220 0.614

11/15/1973 0.176 0.501

02/15/1974 0.144 0.401

05/15/1974 0.147 0.431

08/15/1974 0.149 0.418

11/15/1974 0.139 0.369

02/15/1975 0.128 0.331

05/15/1975 0.131 0.340

08/15/1975 0.140 0.383

11/15/1975 0.131 0.342

02/15/1976 0.139 0.371

05/15/1976 0.148 0.404

08/15/1976 0.166 0.459

11/15/1976 0.158 0.434

02/15/1977 0.168 0.465

05/15/1977 0.166 0.466

08/15/1977 0.162 0.449

11/15/1977 0.162 0.458

02/15/1978 0.138 0.368

05/15/1978 0.148 0.401

08/15/1978 0.139 0.368

11/15/1978 0.150 0.414

02/15/1979 0.134 0.356

05/15/1979 0.155 0.421

08/15/1979 0.147 0.396

11/15/1979 0.145 0.387

02/15/1980 0.139 0.369

05/15/1980 0.140 0.372

08/15/1980 0.131 0.346

11/15/1980 0.140 0.370

02/15/1981 0.139 0.372

05/15/1981 0.138 0.366

08/15/1981 0.147 0.403
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

11/15/1981 0.133 0.348

02/15/1982 0.134 0.352

05/15/1982 0.149 0.407

08/15/1982 0.140 0.375

11/15/1982 0.143 0.387

02/15/1983 0.145 0.399

05/15/1983 0.142 0.385

08/15/1983 0.142 0.390

11/15/1983 0.150 0.414

04/01/1984 0.144 0.377

10/01/1984 0.171 0.495

Figure C-46. 50th-percentile sodium whole-body counting data for 

inhalation intakes of type F material, 1960 to 1984. 

● = used in fit, ● = not used in fit, — = fit

Figure C-47. 50th-percentile sodium whole-body counting data for 

ingestion intakes, 1960 to 1984. 

● = used in fit, ● = not used in fit, — = fit
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Table C-47. Statistical summary of 137Cs measured in 

whole-body counts (nCi), 1960 to 1988. 

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

02/15/1960 6.773 9.620

05/15/1960 6.689 9.347

08/15/1960 5.977 9.466

11/15/1960 5.770 8.302

02/15/1961 4.728 6.858

05/15/1961 4.482 6.764

08/15/1961 3.805 6.006

11/15/1961 3.491 5.184

02/15/1961 3.527 5.563

05/15/1961 3.457 5.055

08/15/1961 4.378 6.502

11/15/1961 5.253 7.678

02/15/1963 6.691 10.362

05/15/1963 6.773 9.620

08/15/1963 6.689 9.347

11/15/1963 5.977 9.466

02/15/1964 5.770 8.302

05/15/1964 4.728 6.858

08/15/1964 4.482 6.764

11/15/1964 3.805 6.006

02/15/1965 3.491 5.184

05/15/1965 3.527 5.563

08/15/1965 3.457 5.055

11/15/1965 4.378 6.502

02/15/1966 5.253 7.678

05/15/1966 6.691 10.362

08/15/1966 8.922 13.998

11/15/1966 11.218 16.663

02/15/1967 13.470 19.952

05/15/1967 15.374 23.543

08/15/1967 17.644 26.857

11/15/1967 18.129 26.767

02/15/1968 17.811 28.334

05/15/1968 15.208 23.287

08/15/1968 15.288 23.190

11/15/1968 13.341 19.881

02/15/1969 11.964 18.931

05/15/1969 10.544 16.593

08/15/1969 8.526 13.837

11/15/1969 8.198 13.210

02/15/1970 6.417 12.039

05/15/1970 5.877 9.856

08/15/1970 5.417 9.242

11/15/1970 4.770 8.025

02/15/1971 4.049 7.580

05/15/1971 3.195 6.012

08/15/1971 2.574 4.939

11/15/1971 2.667 5.467
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

02/15/1972 2.824 5.511

05/15/1972 3.022 5.516

08/15/1972 2.176 4.728

11/15/1972 2.222 4.332

02/15/1973 2.0544 4.5035

05/15/1973 1.8482 3.4682

08/15/1973 1.8296 3.3453

11/15/1973 1.6404 3.3004

02/15/1974 1.2505 3.0204

05/15/1974 0.9186 2.1838

08/15/1974 1.0889 2.4621

11/15/1974 1.1091 2.7017

02/15/1976 1.0488 2.1653

05/15/1976 1.1197 2.6952

08/15/1976 0.9322 2.1818

11/15/1976 1.6176 3.9736

02/15/1977 1.3651 2.9877

05/15/1977 1.4688 3.4256

08/15/1977 1.1219 2.9244

11/15/1977 0.8683 2.2466

02/15/1978 1.0525 2.6430

05/15/1978 0.7876 2.3797

08/15/1978 0.7788 2.2337

11/15/1978 1.0824 2.6808

02/15/1979 1.1775 2.9579

05/15/1979 0.8671 2.0518

08/15/1979 0.9219 2.1979

11/15/1979 1.2393 2.6536

02/15/1980 1.1519 2.8199

05/15/1980 0.8175 2.9124

08/15/1980 0.6695 1.8971

11/15/1980 0.7321 1.9398

02/15/1981 0.8278 2.1018

05/15/1981 0.4992 1.1927

08/15/1981 0.3353 0.9534

11/15/1981 0.4740 1.0964

02/15/1982 0.5655 1.3484

05/15/1982 0.4695 1.2468

08/15/1982 0.4364 1.0269

11/15/1982 0.5316 1.2952

02/15/1983 0.4257 1.1238

05/15/1983 0.3023 0.8154

08/15/1983 0.3705 0.9053

11/15/1983 0.4354 1.1195

02/15/1984 0.2487 0.6537

05/15/1984 0.2461 0.6285

08/15/1984 0.2430 0.6134

11/15/1984 0.2398 0.5927

02/15/1985 0.2450 0.6327

05/15/1985 1.1416 2.7971

08/15/1985 1.1220 2.8167
Document No. ORAUT-TKBS-0006-5 Revision No. 07 Effective Date: 09/01/2020 Page 179 of 180

ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

Effective bioassay 

date

GM 

(50th)

GM × GSD 

(84th)

11/15/1985 1.1235 2.8170

04/01/1986 1.1169 2.7720

10/01/1986 1.1106 2.7455

04/01/1987 0.9186 2.1838

10/01/1987 1.0889 2.4621

04/01/1988 1.1091 2.7017

10/01/1988 1.0488 2.1653

Figure C-48. 50th-percentile 137Cs whole-body counting data for 

inhalation intakes of type F material, 1960 to 1988. 

● = used in fit, ● = not used in fit, — = fit

SRDB REF ID NUMBERS

Table C-48. Hanford monthly reports. 

434, 442, 451, 5044, 5047, 5054, 5055, 5057, 5060, 5072, 5087, 5090, 5098, 5199, 33220, 33268, 

33271, 34248, 34251, 34254, 34255, 34267, 34270, 34271, 34272, 34273, 34281, 34285, 34290, 

34293, 34294, 34302, 34304, 34307, 34308, 34309, 34310, 34341, 34350, 34355, 34604, 35862, 

35863, 35864, 35866, 35867, 36737, 36740, 36744, 36746, 36751, 36754, 36757, 36779, 36780, 

36781, 36794, 36802, 36804, 36829, 36831, 36834, 36837, 36849, 36864, 36871, 36889, 36895, 

36926, 36935, 36943, 37089, 37229, 37234, 37239, 37244, 37246, 37248, 37256, 37261, 37265, 

37270, 37274, 37280, 37281, 37282, 37297, 37311, 37323, 37334, 37341, 37346, 37364, 37374, 

37390, 37400, 37405, 37409, 37411, 37418, 37422, 37435, 37440, 37474, 37556, 37559, 37561, 

37563, 37564, 37568, 37569, 37571, 37580, 37582, 37586, 37587, 37591, 40711, 40712, 40726, 

40761, 40876, 40878, 40879, 40880, 40891, 40893, 40894, 40956, 40957, 40958, 40960, 40963, 

40966, 40968, 40971, 40974, 40975, 40976, 40977, 40978, 40979, 40980, 40981, 40983, 40984, 

40985, 40996, 41005, 41006, 41007, 41008, 41200, 41201, 41202, 41203, 41204, 41205, 41206, 

41207, 41208, 41209, 41210, 41211, 41212, 41213, 41214, 41223, 41224, 41226, 41228, 41229, 

41230, 41231, 41233, 41270, 41271, 41306, 41307, 41311, 41314, 41316, 41319, 41322, 41323, 

41332, 41333, 41334, 41335, 41336, 41337, 41339, 41340, 41341, 41342, 41343, 41344, 41345, 

41346, 41338, 51245, 52690, 52693, 53051, 53364, 53366, 53367

Table C-49. Tritium hard-copy bioassay data. 

54902, 54903, 54905, 54906, 54917, 54924, 54925, 54927, 54928, 54929, 54930, 54932, 54937, 

54943, 54944, 54947, 54948, 54957, 54960, 54961, 54963, 54964, 54967, 54970, 54971, 54972, 

54973, 54974, 54975, 54976, 54977, 54978, 54979, 54980, 54981, 54982, 54983, 54984, 54986, 
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ATTACHMENT C

INTERNAL DOSIMETRY CO-EXPOSURE DATA (continued)

54987, 54988, 54989, 54990, 54991, 54992, 54993, 54994, 54995, 55036, 55038, 55039, 55040, 

55041, 55043, 55049, 55053, 55060, 55071, 55073, 55076, 55079, 55083, 55094, 55095, 55096, 

55097, 55098, 55099, 55100, 55101, 55102, 55103, 55104, 55105, 55106, 55107, 55108, 55109, 

55110, 55111, 55112, 55113, 55114, 55115, 55116, 55117, 55118, 55119, 55139, 55140, 55143, 

55145, 55146, 55147, 55152, 55167, 55168, 55169, 55170, 55172, 55175, 55178, 55184, 55185, 

55186, 55187, 55188, 55189, 55190