ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 163

DOE Review Release 11/03/2010

Document Title:

Hanford Site – Occupational Internal Dose

Document Number: ORAUT-TKBS-0006-5 

Revision: 04 

Effective Date: 10/20/2010

Type of Document: TBD

Supersedes: Revision 03 

Subject Expert(s): Fred Duncan, Matthew Arno and Edward D. Scalsky

Site Expert(s): Donald E. Bihl

Approval: Signature on File Approval Date: 10/07/2010

Fred Duncan, Document Owner

Concurrence: Signature on File Concurrence Date: 10/12/2010

Chris M. Tornes, Objective 1 Representative

Concurrence: Signature on File Concurrence Date: 10/14/2010

Scott R. Siebert, Objective 3 Representative

Concurrence: Signature on File Concurrence Date: 10/11/2010

James P. Griffin, Deputy Project Director

Approval: Signature on File Approval Date: 10/20/2010

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 2 of 163

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: added stated about 244Cm recovery 

in 325 Bldg. Section 5.2.5: added statement about work with 37% 

enriched U; changed period for 4 μg/L reporting level to 1974; 

increased 239Pu contaminant level in RU to 0.8 nCi/gU; 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/gU; added statement about 

work with 37% enriched U; 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. Sec. 5.3.1: 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 3 of 163

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 specificfacilities 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.

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 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 4 of 163

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 DRs concerning uncertainty. 

In Section 5.6, added reference to OTIB-54 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 Duncan.

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 Duncan. 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 5 of 163

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

11

5.1 Introduction ............................................................................................................................13

5.1.1 Purpose......................................................................................................................14

5.1.2 Special Exposure Cohort Petition Information for Hanford.......................................... 14

5.1.3 Scope.........................................................................................................................14

5.2 In Vitro Minimum Detectable Activities, Analytical Methods, and Reporting 

Protocols................................................................................................................................16

5.2.1 Plutonium ...................................................................................................................17

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

5.2.3 Curium........................................................................................................................25

5.2.4 Tritium ........................................................................................................................25

5.2.5 Uranium......................................................................................................................28

5.2.6 Fission Product Analysis.............................................................................................34

5.2.7 Strontium....................................................................................................................35

5.2.8 Promethium................................................................................................................36

5.2.9 Polonium ....................................................................................................................37

5.2.10 Neptunium..................................................................................................................38

5.2.11 Other Limited-Exposure Radionuclides.......................................................................39

5.3 In Vivo Minimum Detectable Activities, Analytical Methods, and Reporting 

Protocols................................................................................................................................40

5.3.1 Whole-Body Counters ................................................................................................40

5.3.2 Chest Counters ..........................................................................................................44

5.3.3 Thyroid Counters........................................................................................................47

5.3.4 Thorium Exposure and Monitoring..............................................................................48

5.3.5 Head Counters and Other Counts ..............................................................................51

5.3.6 General Notes About Items in the Database...............................................................51

5.4 Mixtures ................................................................................................................................. 52

5.4.1 Fission and Activation Products..................................................................................52

5.5 Interferences and Uncertainties.............................................................................................. 53

5.5.1 Contamination of Samples..........................................................................................53

5.5.2 Uncertainties ..............................................................................................................54

5.6 Unmonitored Workers ............................................................................................................ 54

5.6.1 Radionuclides Not Presented in Attachment C ...........................................................55

5.6.2 Assignments for Unmonitored Workers ......................................................................61

5.7 Attributions and Annotations .................................................................................................. 66

References ........................................................................................................................................81

Glossary ............................................................................................................................................92
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 6 of 163

ATTACHMENT A, TOLERANCE DOSE AND TOLERANCE VALUES AT 

HANFORD .......................................................................................................... 98

ATTACHMENT B, CODES USED IN BIOASSAY AND INTERNAL DOSE 

RECORDS ....................................................................................................... 100

ATTACHMENT C, INTERNAL DOSIMETRY COWORKER DATA FOR 

THE HANFORD SITE ....................................................................................... 110

C.1 Purpose ............................................................................................................................... 114

C.2 Overview.............................................................................................................................. 114

C.3 Data ..................................................................................................................................... 117

C.3.1 Selected Bioassay Data............................................................................................117

C.3.1.1 Plutonium Urinalysis................................................................................... 117

C.3.1.2 Uranium Urinalysis ..................................................................................... 119

C.3.1.3 Strontium Urinalysis ................................................................................... 121

C.3.1.4 Promethium Urinalysis ............................................................................... 122

C.3.1.5 Zinc-65 in Whole-Body Counts................................................................... 123

C.3.1.6 Sodium-24 in Whole-Body Counts ............................................................. 124

C.3.1.7 Cesium-137 in Whole-Body Counts ........................................................... 125

C.3.2 Analysis....................................................................................................................126

C.3.3 Tritium Dose Evaluation............................................................................................126

C.4 Intake Modeling.................................................................................................................... 129

C.4.1 Assumptions.............................................................................................................129

C.4.2 Bioassay Fitting ........................................................................................................130

C.4.3 Radionuclides and Material Types............................................................................ 130

C.4.3.1 Plutonium................................................................................................... 130

C.4.3.1.1 Type M ..................................................................................... 130

C.4.3.1.2 Type S...................................................................................... 131

C.4.3.2 Uranium ..................................................................................................... 131

C.4.3.2.1 Type F ...................................................................................... 131

C.4.3.2.2 Type M ..................................................................................... 132

C.4.3.2.3 Type S...................................................................................... 132

C.4.3.2.4 Strontium-90 ............................................................................. 133

C.4.3.3 Promethium-147 ........................................................................................ 133

C.4.3.3.1 Type M ..................................................................................... 133

C.4.3.3.2 Type S...................................................................................... 133

C.4.3.4 Zinc-65....................................................................................................... 133

C.4.3.4.1 Inhalation Intakes...................................................................... 134

C.4.3.4.2 Ingestion Intakes....................................................................... 134

C.4.3.5 Sodium-24.................................................................................................. 134

C.4.3.5.1 Inhalation Intakes...................................................................... 134

C.4.3.5.2 Ingestion Intakes....................................................................... 134

C.4.3.6 Cesium-137................................................................................................ 135

C.5 Assignment of Intakes and Doses ........................................................................................ 135

C.5.1 Plutonium .................................................................................................................137

C.5.2 Uranium....................................................................................................................137

C.5.3 Strontium-90.............................................................................................................137

C.5.4 Promethium-147.......................................................................................................139
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 7 of 163

C.5.5 Zinc-65 .....................................................................................................................139

C.5.6 Sodium-24................................................................................................................140

C.5.7 Cesium-137..............................................................................................................141

C.5.8 Additional Radionuclides ..........................................................................................141

LIST OF TABLES

TABLE TITLE

5-1 Routine plutonium urinalysis detection levels .........................................................................

PAGE

18

5-2 MDAs for nonroutine plutonium excreta analyses .................................................................. 19

5-3 Activity composition of Hanford reference weapons-grade plutonium mixture (6%)................ 20

5-4 Activity composition of Hanford reference fuel-grade plutonium mixture (12%) ...................... 21

5-5 Activity composition of Hanford reference commercial power fuel-grade plutonium

mixture...................................................................................................................................22

5-6 Routine 241Am urinalysis detection levels ............................................................................... 24

5-7 MDAs for nonroutine 241Am excreta analyses......................................................................... 25

5-8 Routine curium urinalysis detection levels.............................................................................. 26

5-9 MDAs for nonroutine curium excreta analyses ....................................................................... 26

5-10 Routine tritium urinalysis detection levels............................................................................... 27

5-11 Radiological characteristics of Hanford uranium mixtures ...................................................... 28

5-12 Impurities in RU at Hanford.................................................................................................... 29

5-13 Inhalation class for Hanford uranium compounds................................................................... 29

5-14 Routine uranium urinalysis detection levels............................................................................ 31

5-15 MDAs for nonroutine uranium excreta analyses ..................................................................... 31

5-16 Highest level of impurities as measured for the three campaigns........................................... 33

5-17 Highest impurity concentrations in 233U .................................................................................. 33

5-18 Routine fission product urinalysis detection levels.................................................................. 34

5-19 Routine 90Sr urinalysis detection levels .................................................................................. 36

5-20 Routine 147Pm urinalysis detection levels ............................................................................... 36

5-21 Routine 210Po urinalysis detection levels ................................................................................ 38

5-22 MDAs for nonroutine 210Po excreta analyses.......................................................................... 38

5-23 Routine whole-body counting detection levels........................................................................ 41

5-24 Mean body burdens of 137Cs from fallout in the United States ................................................ 44

5-25 Routine chest counting detection levels ................................................................................. 46

5-26 Buildings with thorium processing .......................................................................................... 48

5-27 Thorium-232 intakes from whole-body count measurements ................................................. 49

5-28 Ratios of 232Th to 228Ac in the whole body after end of intake period, type M .......................... 50

5-29 Ratios of 232Th to 228Ac in the whole body after end of intake period, type S........................... 50

5-30 Statistical parameters of Hanford recorded tritium doses and associated intakes .................. 56

5-31 Intakes for workers with external monitoring but essentially no bioassay ............................... 61

B-1 Codes and radionuclides associated with bioassay at Hanford ............................................ 100

B-2 Sample type codes............................................................................................................... 103

B-3 Bioassay reason codes ........................................................................................................ 104

B-4 Excreta sample kit codes ..................................................................................................... 105

B-5 In vivo body location codes .................................................................................................. 106

B-6 Excreta unit codes................................................................................................................ 106

B-7 Excreta processing codes .................................................................................................... 106

B-8 Excreta laboratory codes...................................................................................................... 107

B-9 Excreta no-sample codes..................................................................................................... 107
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 8 of 163

B-10 In vivo invalid result codes ................................................................................................... 107

B-11 INTERTRAC mode-of-intake codes ..................................................................................... 108

B-12 INTERTRAC evaluation reason codes ................................................................................. 108

B-13 INTERTRAC source-of-intake codes.................................................................................... 108

B-14 INTERTRAC miscellaneous codes....................................................................................... 108

B-15 Special whole-body count resolution codes.......................................................................... 109

C-1 Time intervals for statistical analysis of plutonium urine samples ......................................... 120

C-2 Time intervals for statistical analysis of uranium urine samples............................................ 121

C-3 Time intervals for statistical analysis of 90Sr urine samples .................................................. 122

C-4 Time intervals for statistical analysis of 147Pm urine samples ............................................... 123

C-5 Time intervals for statistical analysis of 65Zn whole-body counts .......................................... 124

C-6 Time intervals for statistical analysis of 24Na whole-body counts .......................................... 125

C-7 Time intervals for statistical analysis of 137Cs whole-body counts......................................... 125

C-8 Plutonium mixture usage...................................................................................................... 129

C-9 Type M 239+240Pu intake periods and rates ............................................................................ 131

C-10 Type S 239+240Pu intake period and rates .............................................................................. 131

C-11 Type F uranium intake periods and rates ............................................................................. 132

C-12 Type M uranium intake periods and rates ............................................................................ 132

C-13 Type S uranium intake periods and rates ............................................................................. 132

C-14 Type F 90Sr intake periods and rates.................................................................................... 133

C-15 Type M 147Pm intake periods and rates ................................................................................ 133

C-16 Type S 147Pm intake periods and rates................................................................................. 133

C-17 Type S zinc-65 inhalation intake periods and rates .............................................................. 134

C-18 Zinc-65 ingestion intake periods and rates ........................................................................... 134

C-19 Type F 24Na inhalation intake periods and rates................................................................... 135

C-20 Sodium-24 ingestion intake periods and rates...................................................................... 135

C-21 Type F 137Cs intake periods and rates.................................................................................. 135

C-22 Type M plutonium-239 intake periods and rates................................................................... 137

C-23 Type S plutonium-239 intake period and rates ..................................................................... 137

C-24 Uranium Type F intake periods and rates............................................................................. 138

C-25 Uranium Type M intake periods and rates............................................................................ 138

C-26 Uranium Type S intake periods and rates ............................................................................ 138

C-27 Type F 90Sr intake periods and rates.................................................................................... 139

C-28 Promethium-147 intake periods and rates............................................................................ 139

C-29 Zinc-65 intake periods and rates .......................................................................................... 140

C-30 Sodium-24 intake periods and rates..................................................................................... 140

C-31 Type F cesium-137 intake periods and rates........................................................................ 141

C-32 Statistical summary of 239Pu 24-hour urinary excretion rates, 1946 to 1988 ......................... 142

C-33 Statistical summary of uranium 24-hour urinary excretion rates, 1948 to 1988..................... 148

C-34 Statistical summary of 90Sr 24-hour urinary excretion rates, 1966 to 1988 ........................... 152

C-35 Statistical summary of 147Pm 24-hour urinary excretion rates, 1967 to 1977......................... 153

C-36 Statistical summary of 65Zn measured in whole-body counts, 1960 to 1984 ......................... 156

C-37 Statistical summary of 24Na measured in whole-body counts, 1960 to 1984......................... 160

C-38 Statistical summary of 137Cs measured in whole-body counts, 1960 to 1988........................ 162

C-39 Hanford monthly reports....................................................................................................... 163

C-40 Tritium hard-copy bioassay data .......................................................................................... 163
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 9 of 163

LIST OF FIGURES

FIGURE TITLE

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

PAGE

115

C-2 Uranium bioassay data in REX and from Hanford monthly reports....................................... 116

C-3 Tritium bioassay data........................................................................................................... 127

C-4 Individual tritium bioassay or dose data by year................................................................... 128

C-5 Tritium dose from hard-copy records and the main text........................................................ 128

C-6 Production rates in tons of uranium times fuel burnup, 1944 to 1972 ................................... 136

C-7 50th-percentile plutonium urinalysis data for intakes of Type M material, 1946 to 1948 ....... 143

C-8 50th-percentile plutonium urinalysis data for intakes of Type M material, 1949 to 1952 ....... 144

C-9 50th-percentile plutonium urinalysis data for intakes of Type M material, 1953 to 1981 ....... 144

C-10 50th-percentile plutonium urinalysis data for intakes of Type M material, 1982 to 1988 ....... 144

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

based on four independent intakes, compared to bioassay data .......................................... 145

C-12 50th-percentile plutonium urinalysis data for intakes of Type S material, 1946 to 1948........ 145

C-13 50th-percentile plutonium urinalysis data for intakes of Type S material, 1949 to 1952........ 146

C-14 50th-percentile plutonium urinalysis data for intakes of Type S material, 1953 to 1981........ 146

C-15 50th-percentile plutonium urinalysis data for intakes of Type S material, 1982 to 1988........ 146

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

based on four independent intakes, compared to bioassay data .......................................... 147

C-17 50th-percentile uranium urinalysis data for intakes of Type F material, 1948 to 1988........... 149

C-18 50th-percentile uranium urinalysis data for intakes of Type M material, 1948 to 1988.......... 149

C-19 50th-percentile uranium urinalysis data for intakes of Type S material, 1948 to 1952........... 149

C-20 50th-percentile uranium urinalysis data for intakes of Type S material, 1953 to 1956........... 150

C-21 50th-percentile uranium urinalysis data for intakes of Type S material, 1957 to 1961........... 150

C-22 50th-percentile uranium urinalysis data for intakes of Type S material, 1962 to 1974........... 150

C-23 50th-percentile uranium urinalysis data for intakes of Type S material, 1975 to 1983........... 151

C-24 50th-percentile uranium urinalysis data for intakes of Type S material, 1984 to 1988........... 151

C-25 Predicted 50th-percentile urinary excretion of Type S uranium from 1946 to 1988 

based on six independent intakes, compared to bioassay data............................................ 151

C-26 50th-percentile strontium urinalysis data for intakes of Type F material, 1965 to 1988......... 152

C-27 50th-percentile promethium urinalysis data for intakes of Type M material, 1966 to 

1969..................................................................................................................................... 153

C-28 50th-percentile promethium urinalysis data for intakes of Type M material, 1970 to 

1979..................................................................................................................................... 153

C-29 Predicted 50th-percentile urinary excretion of Type M promethium from 1966 to 1979 

based on two independent intakes, compared to bioassay data........................................... 154

C-30 50th-percentile promethium urinalysis data for intakes of Type S material, 1966 to 

1969..................................................................................................................................... 154

C-31 50th-percentile promethium urinalysis data for intakes of Type S material, 1970 to 

1979..................................................................................................................................... 154

C-32 Predicted 50th-percentile urinary excretion of Type S promethium from 1966 to 1979 

based on two independent intakes, compared to bioassay data........................................... 155

C-33 50th-percentile zinc whole-body counting data for inhalation intakes of Type S 

material, 1960 to 1963 ......................................................................................................... 157

C-34 50th-percentile zinc whole-body counting data for inhalation intakes of Type S 

material, 1964 to 1967 ......................................................................................................... 157

C-35 50th-percentile zinc whole-body counting data for inhalation intakes of Type S 

material, 1968 to 1970 ......................................................................................................... 157

C-36 50th-percentile zinc whole-body counting data for inhalation intakes of Type S 

material, 1971 to 1972 ......................................................................................................... 158
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 10 of 163

C-37 50th-percentile zinc whole-body counting data for inhalation intakes of Type S 

material, 1973 to 1977 ......................................................................................................... 158

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

material, 1978 to 1984 ......................................................................................................... 158

C-39 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....................................................................................................................... 159

C-41 50th-percentile sodium whole-body counting data for inhalation intakes of Type F 

material, 1960 to 1984 ......................................................................................................... 161

C-42 50th-percentile sodium whole-body counting data for ingestion intakes, 1960 to 1984......... 161

C-43 50th-percentile 137Cs whole-body counting data for inhalation intakes of Type F 

material, 1960 to 1988 ......................................................................................................... 163
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 11 of 163

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

Bq becquerel

CFR Code of Federal Regulations

Ci curie

cm centimeter

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 diethylenetriaminepentaacetic acid

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

FFTF Fast Flux Test Facility

FR Federal Register

ft feet

g gram

GeLi lithium-drifted germanium

GM geometric mean

GSD geometric standard deviation

HERB Health-Related Energy Research Branch

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

IMBA Integrated Modules for Bioassay Assessment

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt

KPA kinetic phosphorescence analysis

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

m meter

MDA minimum detectable activity or, for elemental uranium, minimum detectable amount
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 12 of 163

MDC minimum detectable concentration

MeV megaelectron-volt

min minute

mL milliliter

MPBB maximum permissible body burden

MPC maximum permissible concentration

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

ORAU Oak Ridge Associated Universities

ORE Occupational Radiological Exposure (database)

pCi picocurie

PNL Pacific Northwest Laboratory

PNNL Pacific Northwest National Laboratory

POC probability of causation

ppb parts per billion

ppm parts per million

PRTR Plutonium Recycle Test Reactor

PUREX Plutonium–Uranium Extraction (facility)

RDA reliably detectable activity

REDOX Reduction-Oxidation (plant)

REX Radiological Exposure (database)

RU recycled uranium

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

SRS Savannah River Site

TBD technical basis document

TPU total propagated uncertainty

TTA thenoyl trifluoroacetone

U.S.C. United States Code

UNC United Nuclear Company

UST United States Testing Company

yr year

μCi microcurie

μg microgram

μm micrometer

§ section or sections
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 13 of 163

5.1 INTRODUCTION

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

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

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

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

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

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

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section 

of EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42 

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

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

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

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

external occupational radiation exposures are considered valid for inclusion in a dose reconstruction. 

No efforts are made to determine the eligibility of any fraction of total measured exposure for inclusion 

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

occupationally derived (NIOSH 2007):

] guidelines established under 

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

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) define “performance of duty” for 

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2007).

• 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

 

1

 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 14 of 163

5.1.1

This TBD documents 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.

Purpose

5.1.2

The current status of Special Exposure Cohort (SEC) petitions for Hanford are: 

Special Exposure Cohort Petition Information for Hanford

Classes Added to the SEC

• Employees of 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 workdays aggregating at least 

250 workdays or in combination with workdays within the parameters established for one or 

more other classes of employees in the SEC (72 FR 55214). 

• Employees of DOE, its predecessor agencies, and DOE contractors or subcontractors who 

worked from September 1, 1946 through December 31, 1961 in the 300 area, or 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 workdays aggregating at least 250 

workdays occurring either solely under this employment or in combination with workdays 

within the parameters established for one or more other classes of employees in the SEC (73 

FR 37459).

• All employees of DOE, its predecessor agencies, and their contractors and subcontractors 

who worked at the Hanford Site in Richland, Washington, from October 1, 1943 through June 

30, 1972, for a number of workdays aggregating at least 250 workdays, occurring either solely 

under this employment or in combination with workdays within the parameters established for 

one or more other classes of employees included in the SEC (NIOSH 2009). 

Dose reconstruction guidance in this document for periods prior to July 1, 1972 is presented to 

provide a technical basis for partial dose reconstructions for non presumptive cancers not covered 

within the SEC class through June 30, 1972. NIOSH found that it was infeasible to reconstruct doses 

for inadequately monitored radionuclides such as polonium, thorium, and neptunium. Although 

NIOSH found that it is not possible to completely reconstruct radiation doses for the proposed class, 

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 employed at the Hanford site during the 

period from October 1, 1943 through June 30, 1972, but who do not qualify for inclusion in the SEC, 

may be performed using these data as appropriate.

5.1.3

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.

Scope
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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).

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. Attachment C, “Internal Dosimetry Coworker Data for the Hanford Site,” provides 

guidance on assigning intakes for the years before bioassay.

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

started in this timeframe 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.

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 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 16 of 163

used for scheduling bioassay but not for reporting results of the bioassay. 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 Xand gamma-ray analysis on an excreta sample using the low-energy photon detector (LEPD: a thinwindow 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.

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.

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].

There is another anomaly in the results from around 1946 to 1950. There is a urinalysis record with 

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

equivalence 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 (PNL), 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 activities (MDAs) for the various analyses reflecting capabilities at the interim 

laboratories; and (2) delays in collection, analysis, and follow-up to high routine results from the 

normal pattern (Lyon et al. 1991, 1992).

5.2.1

As presented in Section 5.1.2, 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 (72 FR 55214). 

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

Plutonium

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 

(autoradiography) started in December 1952. Detection levels for these and subsequent procedures 

are listed in Table 5-1. The definition of detection level no doubt changed over the years, but the 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 18 of 163

levels in Table 5-1 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 

nuclear-track 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 5-1. Routine plutonium urinalysis detection levels (dpm/sample)a

. 

Period MDA Decision level Measured quantity

Before 06/1949 0.96b 0.66 Total Pu alpha

06/1949–11/1952 0.33 Total Pu alpha

12/1952–01/27/1953 0.18 Total Pu alpha

01/28/1953–03/26/1953 0.15 Total Pu alpha

03/27/1953–11/06/1953 0.05 Total Pu alpha

11/07/1953–12/04/1953 0.07 Total Pu alpha

12/1953–04/1955 0.057 Total Pu alpha

05/1955–08/1955 0.027c Total Pu alpha

09/1955–09/1955 0.04c Total Pu alpha

10/1955–09/30/1983 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 Total Pu alpha

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

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

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

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

09/2001–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 [10].

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 [11]. 

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 [12].

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 any 

241Pu or 241Am present in the urine [8].

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. 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). This method of recording was 
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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 [9].

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 are often 

found in the database [15].

The MDAs that are listed 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 5-1 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 Table 5-2.

Table 5-2. MDAs for nonroutine plutonium excreta analyses (dpm/sample) [18]. 

Period

Fecal sample MDA Urine sample MDA

Emergencya Expedite Priority Emergencya Expedite

1/1965–10/1983b 0.9-1.5 NA 0.1-0.15 0.5-0.7 NA

10/1983–1/1985c 9 NA 0.2 0.5 NA

1/1985–6/1990 9 3 0.2 0.5 0.08

6/1990–2/1991d 20 4 NA 2 0.4

2/1991–present 9 3 0.2 0.5 0.08

NA = not available.

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 (BPNL 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 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, probably about 0.2-0.5 dpm/sample considering the state-of-the-art of those 

laboratories.

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 
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(PUREX) facility and the Plutonium Finishing Plant, 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, 

Buschbom, and Sula 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 with an 

acute intake occurring many days prior to 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 [19], plutonium at Hanford was comprised of 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. 

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 [20]. There is evidence at LANL that 

plutonium produced at Hanford in the 1940s had less 240Pu, perhaps closer to 3% (ORAUT 2009), but 

documentation of that at Hanford has not been found. Tables 5-3 and 5-4 list the relative activities of 

plutonium isotopes and 241Am, which grows in from 241Pu, for 6% 240Pu and 12% 240Pu mixtures 

(PNNL 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).

Table 5-3. Activity composition of Hanford reference weapons-grade plutonium mixture (6%). 

Mixture designation:

Fresh (2-

weeks) 5-yr 10-yr 15-yr 20-yr 25-yr 30-yr

Years of aging:a 0.038 5 10 15 20 25 30

Specific activity in mixture (Ci/g)

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

Activity ratios

Pu-239+240:Am-241 1345 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 1507 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 Am-241 from the plutonium mix.

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 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 21 of 163

Table 5-4. Activity composition of Hanford reference fuel-grade plutonium mixture (12%). 

Mixture designation:

Fresh (2-

weeks) 5-yr 10-yr 15-yr 20-yr 25-yr 30-yr

Years of aging:a 0.038 5 10 15 20 25 30

Specific activity in mixture (Ci/g)

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

Activity ratios

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 mix.

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 

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-to238Pu ratio 

from an intake implies weapons grade (e.g., >8), dose reconstructors should use the ratios in Table 5-

3. 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 may be used as a default, favorable to claimant

assumption. Because there is less Am-241 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’s necessary to compare the two.

With regard to the age of the plutonium, until the fifth 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, 20-year old fuelgrade 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 [21]. 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 5-5 provides the activity characteristics of the commercial fuel used in the Nuclear 

Waste Vitrification Project. In addition, the Plutonium Finishing Plant sometimes recycled plutonium 

from other DOE sites, United Kingdom, and commercial reactors.” These materials often had higher 

240Pu and 241 Am content (Hoyt and Teal 2004, pp. 34-38). Plutonium from the West Valley 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 22 of 163

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 5-5. Activity composition of Hanford reference commercial power fuel-grade plutonium mixture 

(PNNL 2003a). 

Mixture designation:

Fresh (2-

weeks) 5-yr 10-yr 15-yr 20-yr 25-yr 30-yr

Years of aging:a 0.038 5 10 15 20 25 30

Specific activity in mixture (Ci/g)

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

Activity ratios

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.

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 post-October-1983 results are separate for 238Pu and 239+240Pu. 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.

Plutonium at Hanford could have existed as absorption type M, S, or a highly insoluble form, Super S

(ORAUT 2008). 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.

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 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 [e.g., T Plant and 

S Plant (also called Reduction-Oxidation (REDOX) Plant in the early years and the PUREX Plant from 

1956)] and at the Plutonium Reclamation Facility (a wing in the Plutonium Finishing Plant). 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 – 

Americium
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 23 of 163

Site Description, for details (ORAUT 2010c). This americium should be treated as americium (rather 

than trace americium atoms bound in a plutonium matrix) [22]. The International Commission on 

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

As presented in Section 5.1.2, 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 (73 FR 37459). 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 1968, 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) (Author unknown 1961).

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 [23]. No units are given, although the magnitude of the results suggest 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 

units are μCi/L is reasonable [24].

One set of four samples for one worker was collected as follow-up 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 Plutonium Finishing Plant [25].

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 [26].

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 six workers who 

were involved in the same potential intake accident in 222-S on May 2, 1967. 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.”
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A memorandum to file from J. J. Jech (1969a), 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 is 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 

(Bihl 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 

(Bihl 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 (Bihl 2003). The MDA in the 1985 statement of work 

(Author unknown 1985) was listed as 0.02 dpm/sample (consistent with the change to alpha 

spectrometry), 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 (Bihl 2003). 

Table 5-6 summarizes what has been uncovered about 241Am MDAs for routine urinalysis.

Table 5-6. Routine 241Am urinalysis detection levels 

(dpm/sample) [27]. 

Period MDA Decision level

1964 4.6E-6 μCi/L Anything detected

1967–6/1969 5.4E-7 μCi/La Anything detected

7/1969–2/1974 2.0 Anything detected

3/1974–10/1983 0.2 Anything detected

10/1983–9/1985 0.04 Anything detected

10/1985–05/1988 0.02 Anything detected

05/1988–06/1990 0.02 0.01

07/1990–10/1991 0.03 0.015

11/1991–4/2000 0.02 0.01

5/2000–8/2001 0.02 Xb + 2.05 × TPUb

9/2001–present 0.02 2 × TPU

a. Not known if MDA or decision level.

b. 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 5-6 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 Table 5-7. These analyses could have been used because of suspected 

intakes of pure 241Am (such as the explosion of an americium exchange column at the Plutonium 

Finishing Plant 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 involve unusual circumstances such as 

using a supposedly sealed source that had ruptured. These intakes were analyzed by urinalysis, so a 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 25 of 163

procedure obviously existed at that time, although it is not part of the contract with UST [28]. On rare

occasions for a serious intake, samples were analyzed for 241Am using a low-energy photon detector 

(LEPD) before any chemistry. This technique came into existence in 1986 or 1987. Its detection level 

was about 5 dpm/sample (Author unknown 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 [29].

Table 5-7. MDAs for nonroutine 241Am excreta analyses (dpm/sample) [30].

Period

Fecal sample MDA Urine sample MDA

Emergencya Expedite Priority Emergencya Expedite

1/1967–2/1974 (b) NA (b) (b) NA

2/1974–1981 (c) NA 4 (c) NA

1982–9/1983 3.6–12 

(3.6 most probable)d

NA 1.2–5.0 

(1.2 most probable)d

0.7–1.0 

(0.7 most probable)d

NA

10/1983–9/1985 200 NA 0.16 1.0 NA

10/1985–6/1989 20 6 0.1 1 0.08

7/1989–10/1991e 20 4 NA 2 0.4

11/1991–present 20 6 0.1 1 0.08

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

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 (BPNL 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.

5.2.3

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 2010c).

Curium

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 (Author unknown 1987). However, the MDAs were 

not always identical with 241Am. Routine urinalysis MDAs for curium are provided in Table 5-8, and 

nonroutine excreta analyses are provided in Table 5-9.

5.2.4

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, “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 byproduct of heavy water reactor operations. Our most recent experience is from operation of the 

Plutonium Recycle Test Reactor (PRTR)” (McConnon 1967). The PRTR started up in 1960 and was 

Tritium
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 26 of 163

Table 5-8. Routine curium urinalysis detection levels 

(dpm/sample) [31]. 

Period MDA Decision level

7/1969–1981 Not specifically mentioned 

1982–9/1983 Listed for emergency processing only

10/1983–4/1988 0.02 Anything detected

5/1988–6/1990 0.02 0.01

6/1990–10/1991 0.03 0.015

11/1991–4/2000 0.02 0.01

5/2000–8/2001 0.02 Xb + 2.05 × TPUa

9/2001–present 0.02 2 × TPU

a. 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 [32].

Table 5-9. MDAs for nonroutine curium excreta analyses (dpm/sample) [33]. 

Period

Fecal samples, MDA Urine samples, MDA

Emergencya Expedite Priority Emergencya Expedite

Before 1982 (b) NA (b) (b) NA

1982–9/1983 10c NA NA 0.5–1.0 (0.5 most probable)c,d NA

10/1983–9/1985 240 NA 0.8 10 NA

10/1985–6/1989 240 70 0.8 1 1.2

7/1989–10/1991e NA NA NA NA NA

11/1991–present 240 70 0.8 1 1.2

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

b. Probably available but MDAs not found.

c. Total alpha; would have included any americium present also.

d. Varied according to sample size over the range shown; the lower value was generally applicable except for very 

large samples (BPNL 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 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.

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, from about 1978 to 2003 (Bihl 2005). PNNL has 

participated in the DOE Tritium Readiness Program in recent years, producing and testing “tritiumproducing burnable absorber rods” and examination of the 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 dates to 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 (Parker 1950a) that referred to the acetylene method 

for urinalysis; it provides a sensitivity of about 1.2 μCi/L in water. 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 lab in the 329 Building (Watters 1955, 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 27 of 163

p. 4). 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.4) Liquid scintillation counting was implemented for tritium bioassay at the Savannah 

River Site (SRS) in 1958, and it is reasonable to expect that Hanford did so at about the same time. 

In the interview with Lardy (Bihl 2003), Lardy stated that liquid scintillation counting of a 1-mL aliquot 

of raw urine has been used since UST was awarded the bioassay contract in 1965.

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 [34]. 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. 3). 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 

1965 statement of work with UST (AEC 1964) shows an MDA of 1 μCi/L (which is consistent with the 

MDA at SRS throughout the 1950s). Table 5-10 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 5-10. Routine tritium urinalysis detection 

levels [35]. 

Period MDA

1949–1960a ≈ 5 μCi/L

1961–1981b 1 μCi/L

1982–10/1991 10 dpm/mL

11/1991–present 20 dpm/mL

a. Dates and MDA 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. G-8). 

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 HT and HTO levels in the facility [36]. The tritium in 

the lithium-aluminum rods was absorbed in the aluminum as a getter; hence, 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 a 

worker(s) had to be exposed to mercury because of the high contamination [37].

Metal tritides were potentially present as part of the Tritium Target Program work starting in 1988, the 

metal mostly likely being zirconium. The irradiated targets were examined and the tritium driven off in 

hot cells. Samples were taken for analysis, but where the analyses were performed has not been 

discovered yet.
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 28 of 163

5.2.5

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 was a radiation work permit for work with “bare, enriched − 37.5% − solid form 

uranium” for the 306 Building for November 9, 1970, through November 9, 1971 (BN 1970; Jech 

1970a). Table 5-11 provides the default uranium mixtures that have been used at the Hanford Site 

since the 1980s (PNNL 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 Dose Reconstruction Project tools are acceptable. In general, personnel working in 

the production facilities (e.g., fuel fabrication, 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 5-12 (DOE 2000). Plutonium-239 can be assumed for the plutonium alpha 

impurity, and 232Th can be assumed for the thorium [38].

Uranium

Table 5-11. Radiological characteristics of Hanford uranium 

mixtures. 

Uranium mixturea,b

Weight percentage Recycled (RU) Commercial fuel

U-234 0.0082 0.0300

U-235 0.9700 2.9600

U-236 0.0680 Negligible

U-238 98.9500 97.0100

Specific constituent activity in mixture (μCi/g, nCi/mg, or pCi/μg)c

U-234 0.5125 1.8750

U-235 0.0210 0.0639

U-236 0.0440 Negligible

U-238 0.3325 0.3260

Total 0.9099 2.2649

Specific constituent activity in mixture (dpm/μg)c

U-234 1.1378 4.1625

U-235 0.0465 0.1419

U-236 0.0977 Negligible

U-238 0.7381 0.7236

Total 2.0200 5.0281

Constituent fraction of total uranium activity in mixture

U-234 0.5632 0.8279

U-235 0.0230 0.0282

U-236 0.0484 Negligible

U-238 0.3654 0.1439

Total 1.0000 1.0000

a. Commercial fuel data from Rich et al. (1988).

b. RU data based on average of data presented by Sula, Carbaugh, and 

Bihl (1991).

c. Can be used to represent specific alpha activity in the mixture as well.
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 29 of 163

Table 5-12. Impurities in RU at Hanford. 

Constituent Maximum alloweda Observed rangeb

Recommended 

levelc

Plutonium 10 ppb U <1–2 ppb U 0.8 nCi Pu-alpha/g U

Neptunium Not established 0.04–0.16 ppm U 0.4 nCi Np-237/g U

Thorium 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

ZrNb-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, Carbaugh, and Bihl 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 [39].

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 5-13. Because the relationship 

between the old lung fluid studies and the ICRP Publication 66 absorption types (ICRP 1994) is not 

established, Table 5-13 also shows approximate ICRP 66 absorption types for intakes from the listed 

facilities. The information from the lung fluid studies indicates that all material types are present 

onsite. Unless there is claim-specific information about the solubility type, 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 5-13. 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)

Uranyl nitrate at PUREX or UO3 Plant F

UCl4 Mf

U carbonate (assumed form after discharge to the soil) Ff,g

a. ICRP (1982).

b. ICRP (1994)

c. Sula, Carbaugh, and Bihl (1991).

d. Because the conversions from the solubility studies to the ICRP (1994) 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) [40].
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 30 of 163

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 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) (Bihl 2003). (Note: 

Lardy 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 

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 PNL workers were monitored by the isotopic analysis 

because of the wide scope of research projects that occurred over the years [41].

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 [42].

Table 5-14 summarizes the routine urinalysis detection levels and Table 5-15 summarizes nonroutine 

detection levels. 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-TBD6000/PNWD-3738, Technical Basis Document: Site Profiles for Atomic Weapons Employers That 

Worked Uranium and Thorium Metals (Battelle 2006).
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 31 of 163

Table 5-14. Routine uranium urinalysis detection levels. 

Period

Elemental (μg/L) Isotopic (dpm/sample)

MDA Decision level MDA Decision level

Before 1948 Not specifically mentioned NA NA

1948–1949 10 Anything detected NA NA

1950–1962a 4 (reporting level) (b) NA NA

1963–8/1964 0.14 (reporting level)c,d

9/1964–1973 4d

1974–1981 0.4 Same as MDA NA NA

1982–9/1983 0.05–0.25 

(0.1 most probable)e

Same as MDA NA NA

10/1983–12/1983 0.03 μg/sample 0.5 μg/samplef 0.035 Same as MDA

1/1984–8/1985 0.03 μg/sample Not known 0.02 Same as MDA

9/1985–6/1990 0.03/0.5 μg/sampleg 0.2 μg/sampleh 0.02 Same as MDA

6/1990–10/1991 0.2/0.5 μg/sampleg 0.2h 0.03 0.15/0.015i

11/1991–present 0.06/0.5 μg/sampleg 0.2f 0.02 0.15/0.010g

NA = not available.

a. 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.

b. 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.

c. GE (1963, p. G-8); GE (1964, p. G-9). 

d. 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.

e. MDAs were based on sample size, but 0.1 μg/L was applied to most sample sizes (BPNL 1982). The database 

shows a change to reporting in μg/sample on July 2, 1982.

f. Values below this were recorded but not followed up as occupational intakes.

g. 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 [43]. 

h. Based on upper level for natural background excretion. See text for discussion.

i. 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 PNNL 2003b). 

Table 5-15. MDAs for nonroutine uranium excreta analyses [44]. 

Period

Analysis 

type

Fecal sample MDA Urine sample MDA 

Emergencya Expedite Priority Emergencya Expedite

1980b Elemental NA NA NA 10 μg/L NA

10/1983–

9/1985

Elemental 8 μg/sample NA 0.3 μg/sample 7 μg/sample NA

Isotopicc 12 dpm/sample NA 0.3 dpm/sample 1 dpm/sample NA

10/1985–

present 

Elemental 8 μg/sample 5 μg/sample 0.3 μg/sample 7 μg/sample 0.5 μg/sample

Isotopic 12 dpm/sample 4 dpm/sample 0.3 dpm/sample 1 dpm/sample 0.12 dpm/sample

6/1990–

10/1991d

Elemental 20 μg/sample 4 μg/sample NA 20 μg/sample 4 μg/sample

Isotopic 20 dpm/sample 4 dpm/sample NA 2 dpm/sample 0.4 dpm/sample

11/1991–

present

Elemental 8 μg/sample 5 μg/sample 0.3 μg/sample 7 μg/sample 0.5 μg/sample

Isotopic 12 dpm/sample 4 dpm/sample 0.3 dpm/sample 1 dpm/sample 0.12 dpm/sample

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

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, 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.

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 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 32 of 163

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 (PNNL 2003a; MacLellan, Wyse, and Scott 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-to238U 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 (PNNL 2003a).

Therefore, 0.2 μg/d was established in 1985 and continues to be used at present by the Site 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 screen 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 are present, dose reconstructors should consider any 

detected uranium as 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.

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 5.3.4 of the Hanford Site Description (ORAUT 2010c) 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. 3–4; Hopkins 1965, 

p. 68), 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 

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 [45]. 

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).
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 33 of 163

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 5-16.

Table 5-16. 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/μgU 0.002 1.8E-06d NG

Th μg/μgU 0.094 2.2E-04 NG

Np μg/μgU NG 2.8E-05 NG

Zr/Nb-95 pCi/μgU NG 0.35 NG

Ru-103/Ru-106 pCi/μgU NG 0.046 NG

Pa-233 pCi/μgU 102 4.4 NG

Th-238 pCi/μgU NG NG 0.84

NG = not given

a. As measured at Z Plant (Fullam 1965, p. 22)

b. Isochem (1967, p. VII-7).

c. Jackson and Walser (1977, p. 175).

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 5-17. 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 5-17. Highest impurity concentrations in 233U (pCi/μgU). 

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-238 8.4E-01

Work with 233U is covered by the SEC. If dose reconstruction is needed, see Section 5.6 for guidance 

on intakes of 233U for unmonitored workers.

There was exposure to highly enriched uranium (HEU) at Hanford, starting in July 1948, occurring 

heavily 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 2010c) for details. The workers exposed to HEU were most likely also 

exposed to natural or slightly enriched uranium. There does not seem to be a way to determine from 

measurements in workers’ files whether 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. The proposed SEC ends June 30, 1972 (NIOSH 2009). If claim-specific evidence is found for 

HEU from July 1, 1972 through October, 1983, the elemental uranium urinalysis and chest counts will 

be used to reconstruct dose. From October 1983 on, isotopic urinalysis results that indicate HEU 

(

235U) will be used to reconstruct dose. 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 34 of 163

5.2.6

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. 

Fission Product Analysis

As presented in Section 5.1.2, 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 (72 FR 

55214). 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 compilation of procedures that is referred to as the Old Bioassay Bible (Author 

unknown 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 (Author unknown 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 5-18 summarizes 

the most accurate detection levels for the fission product urinalysis found to date.

Table 5-18. Routine fission product urinalysis 

detection levels (dpm/sample) [46]. 

Period MDA

1948–2/1956 60a

3/1956–10/1964 70b

1965–approx. 1969 31

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 

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 

2007a). Because the fission product urinalysis did not account for radiocesium, 90Sr should be used 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 35 of 163

as the indicator radionuclide in ORAUT 2007a. 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 [47].

5.2.7

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 (Author unknown 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 is 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 is 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 letter 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 (ERDA 1975). 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.

Strontium

Beginning in 1982, actual analytical results were recorded in the database; that is, results were not 

truncated at a reporting level.

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 [48]. 

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 5-19 summarizes the routine 

urinalysis detection levels for 90Sr procedure.

Dose reconstructors should consider all strontium at Hanford to be absorption type F [50]. It is 

favorable to claimants to assume that 90Sr and total radiostrontium results are 90Sr even though 89Sr 

might be present.
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Table 5-19. Routine 90Sr urinalysis detection levels [49]. 

Period MDA or MDC

Before 1965 Might have been available but MDA not known

1965–1969 1.67E-05 μCi/L (37 dpm/L)

1970–1974 1.00E-06 μCi/L (2.2 dpm/L)

1975–3/1979 50 dpm/samplea

4/1979–1981 5 dpm/sample

1982–9/1983 2.5 dpm/sample

10/1983–6/1990 2 dpm/sample

9/1990–11/1991 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.

5.2.8

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 2010c). There were between 500 to 

600 bioassay samples 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, 

Nelson, and Crook 1969). Twelve volunteers received an injection of 0.1 μCi, and two volunteers 

received an ingestion of 10 μCi. Another study on the behavior of promethium used swine with 

148mPm as the surrogate for 147Pm (McConnon, Cole, and Smith ca. 1971).

Promethium

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 (1995), 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 5-20 lists 

the 147Pm MDAs or minimum detectable concentrations (MDCs) at various times.

Table 5-20. Routine 147Pm urinalysis detection levels [51]. 

Period MDA or MDC

1963–1964 2.00E-06 μCi/250 mLa

1965–3/25/1970 1.67E-05 μCi/L (37 dpm/L)

3/31/1970–1974 1.00E-05 μCi/L (22 dpm/L)

1975–1979 50 dpm/sampleb

1980–1981 20 dpm/sample

1982–9/1983 5 dpm/sample

10/1983–6/1990 4 dpm/sample

11/1991–presentc 30 dpm/sample

a. Ludwick (1964, pp. 3.21–3.24)

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.
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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 [52]. 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 [53].

The first promethium urinalysis development was in 1962 (Perkins 1963a,b, pp. 3.9–3.11). 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 1963a). 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. 3.21-3.24). This method was described as being more amenable to routine processing. The 

stated detection level was 2 x 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 x 10-5 μCi/L.

The one description of the routine procedure that was used by UST appeared in documents from 

1970, 1974 (Lardy 1970; 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. 4.7) but was not used extensively [54].

5.2.9

As presented in Section 5.1.2, a class was recommended by NIOSH for addition 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 2010c) contains a historical summary of work with polonium.

Polonium

A urinalysis procedure was developed in 1948 apparently in response to an accident involving a PoBe source in which two 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. 4).

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.
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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 5-21 and for nonroutine excreta bioassay in Table 5-22.

Table 5-21. Routine 210Po urinalysis detection levels [55]. 

Period MDA or MDC

3/1968–1970 5.4E-7 μCi/L

1971–1974 5.0E-7 μCi/L

1975–1979 1 dpm/samplea

1980–9/1983 0.1 dpm/samplea

10/1983–present No longer listed in the contract except for 

expedited or emergency samples. Probably 

not used.

a. Based on an unusual definition of “analytical limit” and probably 

conservative on the high side. Reporting level listed as 0.5 

dpm/sample.

Table 5-22. MDAs for nonroutine 210Po excreta analyses (dpm/sample unless otherwise noted)[56]. 

Period

Fecal sample MDA Urine sample MDA

Emergencya Expedite Priority Emergencya Expedite

3/1968–1973 NA NA 5.4E-7 μCi/kg NA NA

1974–9/1983 NA NA (b) NA NA

10/1983–9/1985 340 NA NA 0.8 NA

10/1985–6/1989 340 100 NA 0.8 0.1

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

b. Probably available but not listed in the contract.

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

As presented in Section 5.1.2, a class was recommended by NIOSH for addition 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). 

Neptunium

At 222-S (1958 through 1962) and PUREX (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; however, 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 2010c).

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 that was 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. 4).
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The database contains only four bioassay samples for 237Np in the 1970s, all 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 [57].

A few sample results appear in the database in the 1980s, mostly for PNNL employees. The units are 

in disintegrations per minute per sample [58]. 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

Hanford has always been a center for research, first as part of Hanford Works, then (1965 to present) 

as part of 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.

Other Limited-Exposure Radionuclides

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 (Author unknown 1987; Lyon et al. 1991; Carbaugh 1997). 

Baseline samples were obtained from each worker because natural excretion levels had not been 

established. ICRP (1995, 1996) assigns 14C in organic compounds to class SR2.

Carbon-14

Thorium urinalysis capability apparently existed at least as far back as 1955 (Nickelson et al. 1955, 

pp. 6d, 7d). 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 

was 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).

Thorium

There are some urinalysis results for 35S in 1967. According to Lardy (1970), analysis was 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.

Sulphur-35

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 [59].

Radon
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This radionuclide was used for biological tracer studies and, according to one retired researcher, 

“pipetting was done by mouth in the old days” [60]. 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 circa 1965, and there 

might be personnel count results in the records (Keene 1965, p. F-3).

Phosphorous-32

In addition, some unusual radionuclides were isolated in the 325 Building for nuclear medicine studies 

in the mid to late 1990s.

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 halflife, but the need to perform chest counting never arose (Carbaugh 1990; Bauman 1996).

Yttrium-90

Another project circa 1995-96 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.

Thorium- and Actinium-227

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 (2001).] The unit for all in vivo

measurements is nanocuries.

5.3.1

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, McCall, and Palmer 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, same-size NaI detector was added in 1963 (Brady 1964). According to 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 MDAs as those in Mr. Palmer’s Laboratory 

Record Notebook in 1960 [61], so the difference between the systems was apparently not great 

enough to warrant republishing the MDAs. Therefore, the MDAs in Table 5-23 are the only MDAs 

found for the 1960s and 1970s, and apparently were meant to apply generally to the various wholebody 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.

Whole-Body Counters

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 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 41 of 163

Table 5-23. Routine whole-body counting detection levels (nCi).

a

Period Nuclide MDA Decision level Reporting level

1960–1976b

Na-22 1.0 10

Na-24 0.3 0.3

Cr-51 50 50

Fe-59 2.0 10

Co-60 0.4 10

I-131 0.5c 10

Cs-137 0.5 0.5

Zn-65 1.6 0.8

1977–1984d

Na-22 1.0 10

Na-24 0.5 0.31

Cr-51 15 15

Mn-54 2.0 10

Fe-59 4.0 10

Co-60 2.0 10

Zn-65 3.0 0.75

Zr/Nb-95 2.0 10

Ag-110m 2.0 10

Ru-106 12 12

Sb-125 3.0 10

I-131 4.0c 10

Cs-137 2 0.66e

Ce-144 100 100

1985–86

Na-22 1 1.5

Na-24 1 0.31

Mn-54 1 3

Fe-59 2 6

Co-60 1 1

Zn-65 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

1987f 

Na-22 1.5 Anything detected

Mn-54 3 6

Co-60 3 5

Fe-59 6 Anything detected

Cs-137 3 6

No changes for other radionuclides. Anything detected is reported.

1992g New formalism for decision level calculation; “limit” in database changed from MDA to 

decision level.

1993h Actual values, regardless of amount, reported for Co-60 and Cs-137, including negative 

numbers.

1995–10/1999i

Co-60 4 Every result

Cs-137 4 Every result

I-131 5 Every result

Mn-54 3 Every result

Na-22 2 Every result

Na-24 1 Every result

Pr-144 (Ce-144) 230 Every result

Other radionuclides Anything detected
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Period Nuclide MDA Decision level Reporting level

10/1999– 

presenti,j

Co-60 1.25 g Every result

Cs-137 1.3 g Every result

Eu-154 3.75 g Every result

Other radionuclides 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 [62]. 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 [63].

f. From Lyon et al. (1988).

g. From Lyon et al. (1993).

h. [64].

i. Least sensitive of many options during the period. Much better sensitivities were available using the hyperpure 

germanium (HPGe) system (PNL 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).

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-in. by 4-in. 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. 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. Count time is 200 seconds (Palmer, 

Rieksts, and Spitz ca. 1984).

A coaxial HPGe scanning array developed in July 1989 is still in operation 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 
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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 a count on the 

“standup” counter had 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 

started 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 a 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 208

Tl were and are present in the background spectra 

from the shielded rooms used for in vivo measurements of workers at Hanford. 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 the detectors being 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, activity levels of 214Bi were measured periodically 

that were higher than the 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-in. by 16-in. by 

4-in. thick NaI detectors plus one 4-in. by 8-in. 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 use 

of this facility was infrequent, and it was discontinued in August 1995 (Lyon et al. 1996).

From 1960 to 1983, four radionuclides were reported routinely: 24Na, 40K, 65Zn, 137Cs [65]. 

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 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 that were statistically different from 

the general background in Hanford workers and had to come from 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 been occurred before 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) [66].
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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 [67].

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: [68]

• 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 5-24 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 5-24, the dose 

reconstructor may assume the 137Cs result is due to fallout.

Table 5-24. Mean body burdens of 137Cs from fallout in 

the United States (nCi).a

Year Body burden Year Body burden

1953 0.27 1966 9.7

1954 1.1 1967 5.6

1955 2.2 1968 3.5

1956 4.3 1969 2.7

1957 5.1 1970 2.7

1958 6.5 1971 2.7

1959 8.1 1972 2.7

1960 6.8 1973 2.7

1961 4.6 1974 1.6

1962 6.0 1975 1.1

1963 11 1976 1.6

1964 19 1977 1.1

1965 16

a. From NCRP (1988).
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 45 of 163

5.3.2

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. 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. 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 5-25.) 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 [69].

Chest Counters

A solid-state germanium counting system using three planar HPGe detectors (soon upgraded to six) 

replaced the NaI detector in the Iron Room chair counter in 1983 (Palmer, Rieksts, and Spitz ca. 

1984). The HPGe detectors provided better spectral resolution than the NaI detector, thus lower 

backgrounds in the region of interest and better discrimination against radon decay products and 

better detection of low-energy photon emitters in the presence of large activities of high-energy 

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. Counting 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 discernable by naked-eye inspection of the spectrum (Carbaugh et al. 1988). Special 

chest counts as follow-up to high routine chest counts or on special request were twice the normal 

counting time, so the MDAs were somewhat lower.

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, 

Rieksts, and Spitz 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. 

Counting 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 

counting time was increased to 3,000 seconds for the larger area arrays in November 1995; special 

counts and recounts were 3,600 seconds.
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 46 of 163

Table 5-25. Routine chest counting detection levels (nCi)

(Andersen 1971; Lyon et al. 1988, 1989, 1990, 1993, 1994; 

MacLellan et al. 1999; Lynch et al. 2000).

Period Radionuclide MDA

1967 Am-241 0.33

U-238 (Th-234) 6.7

U-235 0.15

1968–1983 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

1983–1986 Am-241 0.24b

U-238 1.1b

U-235 0.08b

Ce-144 0.78b

Eu-154 0.07

1987 Am-241 0.28c

U-238 1.8

U-235 0.12

Ce-144 0.6

Eu-154 0.07

1988–6/1989 Am-241 0.28c

U-238 1.8

U-235 0.12

Ce-144 0.6

Eu-154 0.07

7/1989–1991d Am-241 0.18c

U-238 1.2

U-235 0.08

1992–5/1996e Am-241 0.18c

U-238 3

U-235 0.2

6/1996–10/1999 Am-241 0.28c

U-238 1.6

U-235 0.095

11/1999–present 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 95thpercentile male chest wall (0.2/0.13) (Lynch 2003)

c. Adjusted for 95th-percentile male chest wall.

d. Cerium-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.

Ultrasonic measurements 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 [70].
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 47 of 163

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 [71]. 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 

dependent on how long it had taken the material to go through the separation process and be 

transported 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) [72].

5.3.3

Thyroid counting appears to have started on a limited basis for high-risk workers at least as early as 

1956 (see Section 5.6 for a discussion on thyroid counting in 1945 and 1946). Wilson (1960) states, 

Thyroid Counters

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 (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 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 [73].

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 [74].

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 POC.

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 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 48 of 163

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-sec count.

Thyroid counting for either of the iodine isotopes has been rare since 1987.

5.3.4

As presented in Section 5.1.2, a class was recommended by NIOSH for addition 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).

Thorium Exposure and Monitoring

Thorium exposure occurred at many locations throughout Hanford’s history; the Hanford Site 

Description contains details (ORAUT 2010c). Table 5-26 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. 6d, 7d), 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.

Table 5-26. 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

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 whole-body counting is the bioassay method (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 [75]. The 230Th specification for the thorium oxide was <1 ppm (Author unknown 

1968), which means that 230Th contamination would have contributed <6% of the total alpha activity 

and can be ignored.
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 49 of 163

Routine monitoring for thorium intakes was by whole-body counting based on 228Ac. Note: Use of 

whole-body counting for thorium monitoring was apparently different than at other AEC sites 

that used chest counting. Be sure to recognize that the following discussion and Tables 5-26 

to 5-28 apply only to whole-body counts. Urinalyses were used for special investigations. Jech 

(1967) stated the detection limit for 232Th was 1.2 nCi. Jech (1969b) 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 [76]. 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 

[77]. 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-yr aged mixture, the factors in Table 5-27 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.

Table 5-27. Thorium-232 intakes from whole-body count measurements [78].

Period of chronic 

intake before wholebody 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 wholebody count (pCi/d)

Type M Type S Type M Type S

182 23.17 11.33 1,810 1,310

364 26.56 9.218 1,150 648

546 29.03 7.834 874 412

728 30.40 6.898 704 297

910 30.73 6.234 580 232

1,092 30.28 5.743 483 191

1,274 29.34 5.370 405 162

1,463 28.09 5.068 341 142

1,645 26.79 4.837 291 127

1,827 25.49 4.649 251 115

2,009 24.24 4.494 218 106

2,184 23.10 4.369 192 98.6

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 5-28 and 5-29 provide the 

ratios of 232Th to 228Ac 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 1960s Jech letters (1967, 1969a) 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 are 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 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 50 of 163

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 [79].

Table 5-28. Ratios of 232Th to 228Ac in the whole body after end of intake 

period, type M [80].

Years after end 

of chronic intake

Period of chronic intake (yr)/Ratio

1 2 3 4 5 6

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 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

Table 5-29. Ratios of 232Th to 228Ac in the whole body after end of intake 

period, type S [81].

Years after end of 

chronic intake

Period of chronic intake (yr)/Ratio

1 2 3 4 5 6

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
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 51 of 163

Years after end of 

chronic intake

Period of chronic intake (yr)/Ratio

1 2 3 4 5 6

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

5.3.5

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 [83].

Head Counters and Other Counts

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 [84].

Routine head counting for 90Sr or 147Pm did occur for a while in the 1970s. These were not very 

sensitive and there is the question about what a head count means in relation to the activity in the 

total skeleton. The same worker will hopefully also have 90Sr urinalysis results. The latter should be 

given preference for confirming or quantifying an intake.

5.3.6

All in vivo results appear to be given in nCi [85]. “Limits” were MDAs, which were treated the same as 

decision levels until 1992. The decision level is listed under “limits” starting in January 1992 [86].

General Notes About Items in the Database

Sometimes a radionuclide is listed without a value or limit. This 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 [87]. 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 

MDAs) were routinely reported, and in late 1987, 60Co was added [88]. 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 [89]. During this time, if a worker had a count 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 52 of 163

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 [90]. 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) [91]. 

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.

Reporting of radionuclides at levels below the MDA or decision level 

should not be interpreted as implying exposure to those radionuclides.

5.4 MIXTURES

Except in a few facilities in the weapons production cycle (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

As presented in Section 5.1.2, 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 (72 FR 

55214). Therefore, dose reconstructions will not include any doses from fission products before 

August 31, 1946. 

Fission and Activation Products

For fission and activation product mixtures through 1988 (1 year after N Reactor shut down), 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 2007a or latest revision). 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. ORAUT-OTIB-0054 includes guidance on intake mixtures that applies to workers in 

different types of work or facilities.

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 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 
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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 (2007a). These 

were 65Zn for essentially all workers because of its presence in the Richland drinking water through 

about 1972, 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 Fast Flux Test Facility (FFTF) for 1980 through 1994 

[92]. 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 5-23. 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 5-23 for all years. When they are not listed in Table 5-23, dose 

reconstructors should assume the MDA is 3 times the MDA for 60Co for 1964 through 1984 [93].

ORAUT (2007a) 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 ratios relative to fission products such 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 Super S) 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 [94].

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

Home collection of excreta samples started very early in the bioassay program (Wilson 1987); 

therefore, 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.

Contamination of Samples

It is likely that a contaminated sample shows as an obvious outlier in the dataset for a given worker.

For in vivo measurements, contamination can occur as external to the body or, in the case of chest 

counting, as external to the lung. If a follow-up in vivo count on the same day or within a few days 
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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 for the bioassay measurements were included in the database starting in late 1981 for 

excreta measurements [95]. 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 1987). 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.

Uncertainties

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 

evening-through-morning periods. Medley, Kathren, and Miller (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 [96]. 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 [97]. 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 [98]. Uncertainty in reproducibly positioning a person to get 

the same result was studied at Hanford and found to be about 5% [99]. 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) [100].

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 (PNNL 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 [101]. 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.

5.6 UNMONITORED WORKERS

For unmonitored workers with external dosimetry, dose reconstructors should use the intakes from the 

Attachment C coworker analysis in combination with the special instructions in this section. When 

assigning intakes of fission products, intakes of associated radionuclides as discussed in ORAUT-
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OTIB-0054 (ORAUT 2007a or latest revision) 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 [102]. 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/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 

[103]. 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 coworker intakes in Attachment C do not address radioiodines, tritium, or 233U. 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

Guidance for assignment of intakes of radioiodines, tritium, 233U, HEU, 210Po, 241Am, 244Cm, 237Np, and 

99Tc by unmonitored workers is provided in this section. 

Radionuclides Not Presented in Attachment C

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. 6; Parker 1950b, p. 6; Parker 1951, p. 6; Parker 1952, p. 8). 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 

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 [104]. From these two datasets, the 

assumption was made that was favorable to claimants that at-risk workers had thyroid burdens less 
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than 30 nCi, which would result from a daily intake of 20,000 pCi [105]. 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 whole-body 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 [106]. 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 [107].

Tritium

Tritium bioassay results were not included in the database until 1984 [108]. 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 [109]. 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 

Plutonium Recycle Test Reactor 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 5-30 should be applied.

Table 5-30. Statistical parameters of Hanford recorded tritium doses and associated intakes [112].

Year GM (mrem) GSD

GM doses normalized to 

present model (mrem)

Daily intakes resulting in the 

normalized doses ( × 106 pCi/d)a

1949 29.9 2.9 (3.0)b 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

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

1972 58.9 2.3 (3.0) 55.6 2.3

a. Use as injection mode in IMBA or Chronic Annual Dose Workbook.

b. A default GSD of 3.0 should be substituted for a calculated GSD of less than 3.

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 
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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 [110], 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 certainly the tritium in the latter facilities would 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 [111]. 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 coworker 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 5-30.

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 1998), 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 5-30 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 Plutonium 

Recycle Test Reactor 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 5-30 [113]. The GSDs in Table 5-30 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), 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 
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in the fifth column in Table 5-30). 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 [114] 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 

coworker analysis was performed on these data [115]. 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 [116]. 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 2010c). The 1986 value can be applied to 

workers at those facilities through their respective shutdown dates [117].

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.

Drinking water for workers in the Hanford 400 Area, of which the principal facility was the FFTF, 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 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 (Bihl 2005). 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 (Bihl 2005). However, the annual doses from the post-1985 

intakes are less than 1 mrem and can be ignored [118].

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 maximum permissible concentration for 239Pu and the total beta measurements were 

compared to the maximum permissible concentration 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 plutonium coworker intakes (Attachment C) to 

unmonitored intakes of 241Am, curium, 210Po, and 237Np. Thorium intakes should be based on 

coworker intakes for uranium. However, these guidelines do not apply to all times. Additionally, there 
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are situations when the 95th percentile of the coworker 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 

95th -percentile intake = 50th percentile intake × GSD1.645. 

Specifics are given for each situation below.

Uranium-233

Operators, radiation monitors, and equipment maintenance workers at PUREX who were involved in 

the 233U campaigns might have had intakes without specific monitoring for 233U. The same might be 

true for chemists, chemical technicians, and radiation monitors associated with the Plutonium 

Chemistry Laboratory in Z-Plant in 1965. The highest urinalysis result (4.39 pCi/d) from a batch of 

1970 PUREX worker records was used to estimate potential chronic intakes for the period from 1965 

through 1970 [119]. Because the material was absorption type F, urinary excretion equilibrates 

quickly and the same daily intake can be applied for 1 to 5 years of chronic exposure. The daily 

intake that produced 4.39 pCi/d uranium in urine was 16 pCi [120]. The end of Section 5.2.5 lists 

contaminants that must be added to this intake. If this intake is applied to 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. 

Highly Enriched Uranium (HEU)

Hanford canned and irradiated HEU fuel elements called C and J slugs from 1949 to 1956. 

Numerous, small-scale, special projects with HEU, often alloyed with other metals, occurred at 

various locations in the 1940s, 50s, and 60s. See the Site Description for details. Exposure to HEU 

prior to July 1, 1972 is covered by the SEC, and intakes will not be assigned for HEU for this period 

for partial dose reconstructions for unmonitored workers. No work with unsealed HEU between 1972 

and 1983 (when isotopic uranium urinalyses began) has been discovered. However, small-scale 

projects cannot be ruled out. Intakes will not be assigned for HEU for the period between July 1, 1972 

and October 1983 for partial dose reconstructions for unmonitored workers. However, if there is 

evidence of work with HEU between July 1, 1972 and October 1983, and the worker was monitored 

with routine chest counts and elemental uranium urinalysis even though not specifically monitored for 

HEU, results should be interpreted using HEU as the source of intake and intakes and doses can be 

calculated. 

Thorium

Thorium fuel elements or target elements were produced at Hanford at various times for various 

reasons since 1945. Many facilities handled thorium at one time or another. See Table 5-26 and the 

Site Description for details. Major production of thoria target elements occurred in the 3732 and 3722 

Buildings in 1965 through 1970. Dissolution of the target elements and separation of the thorium and 

233U occurred in three campaigns at PUREX in 1965, 1966, and 1970. After the 1970 campaign 

PUREX thorium was stored in the WR Vault in the 200 West Area and eventually shipped offsite

(mostly to the Fernald Feed Materials Production Center) in the 1970s. Whole body counting was the 

principal method of routine bioassay beginning in 1964 with a few counts in 1962 and 1963. Chest 

counting for thorium replaced whole body counting in 1979. Several workers had thorium urinalyses 

in 1969 and 1970. Exposure prior to July 1, 1972 is covered by the SEC because it was determined 

to be infeasible to reconstruct doses. Only five whole body counts specific for 232Th were conducted 

from 1970 through 1978 and there were a few urinalyses for “total actinides” during this period, but it 

is possible that workers involved with shipping of the thorium from the WR Vault (primarily storage 

and transfer of liquid thorium nitrate solution) or involved with decontamination of the 3722 and 3732 
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Buildings did not receive bioassay monitoring. If there is evidence of exposure to thorium from 

shipping of the material in the WR Vault or decontamination of the 3722 or 2732 Buildings without 

bioassay, unmonitored workers should be assigned intakes of thorium (with progeny per Section 

5.3.4) using the 95th percentile of the uranium coworker intakes (constant distribution). 

Polonium-210

Exposure to 210Po prior to 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 coworker 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 for details. 

Exposure prior to 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 coworker intakes using the 50th percentile 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 for details. Exposure prior to July 1, 1972 

is covered by the SEC, and intakes will not be assigned for this period for partial dose reconstructions 

for unmonitored workers. Incident follow-up 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 

coworker intakes (constant distribution) prior to 1974 and at the 50th percentile (with GSDs) from 

1974 on. 

Neptunium-237

Neptunium-237 was extracted and purified from fuel from 1958 until PUREX shutdown in 1972, with 

small-scale flowsheet development starting in November 1956. Neptunium target elements were 

fabricated, irradiated, and dissolved from 1959 through 1961 and 1966 through 1970. Many facilities 

were involved at different times. See the Site Description for details. Exposure prior to July 1, 1972 is 

covered by the SEC because it was determined to be infeasible to reconstruct doses. Excreta 

bioassay for neptunium was available from 1960 but was not used routinely until the 1980s. No 

evidence of significant exposure to neptunium after 1972 has been found, but small research projects 

with the material can not be ruled out. If an intake estimate is required from July 1, 1972 on, 

unmonitored workers should be assigned intakes at the 95th percentile of the plutonium coworker 

intakes (constant distribution) prior to 1984 and at the 50th percentile (with GSDs) from 1984 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 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 61 of 163

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 REDOX. 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 prior to July 1, 1972 is 

covered by the SEC, and intakes will not be assigned for this period for partial dose reconstructions 

for unmonitored workers. 

5.6.2

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 below. Dose reconstructors should 

check the Site Description for years of operation of specific facilities. 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 the years 1949-55. 1961-69, and 1972, note that it is the tritium 

doses in Table 5-31 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 5-31. 

Assignments for Unmonitored Workers

Table 5-31. Intakes for workers with external monitoring but essentially no bioassay (pCi/d) [121]. 

Radioactive 

material Period 

Daily intake 

(pCi/d)/dose–tritium 

only (mrem/yr) Distribution

Pu September 1, 

1946–present

See Attachment C See Attachment C

U Through present See Attachment C See Attachment C

Sr-90 September 1, 

1946–present

See Attachment C See Attachment C

Pm-147 See Attachment 

C

See Attachment C See Attachment C

Zn-65 September 1, 

1946–present

See Attachment C See Attachment C

Na-24 September 1, 

1946–present

See Attachment C See Attachment C

Cs-137 September 1, 

1946–present

See Attachment C See Attachment C

Other 

fission/activation 

products

September 1, 

1946–present

See ORAUT-OTIB0054 (ORAUT 

2007a)

See ORAUT-OTIB0054 (ORAUT 2007a) 

I-131 (vapor) 1944–1959 2 E+04 Lognormal, GSD =3

I-131 (vapor) 1960–1987 3.4 E+03 Lognormal, GSD =3 N Reactor only 1973 

through 1983.

1-131 (vapor) 1988–present 0

Tritium (HTO) 1944–1948 6.6 E+06 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

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 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 62 of 163

Radioactive 

material Period 

Daily intake 

(pCi/d)/dose–tritium 

only (mrem/yr) Distribution

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 1.3 E+06 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

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 3.2 E+05 Lognormal, GSD = 4 Reactors, fuel 

dissolution

Tritium (HTO) 1972 2.3E+6/55.6 Lognormal, GSD = 3 N Reactor, PUREX 

plant 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 63 of 163

Radioactive 

material Period 

Daily intake 

(pCi/d)/dose–tritium 

only (mrem/yr) Distribution

Tritium (HTO) 1973–1983 3.2 E+05 Lognormal, GSD =4 N Reactor

Tritium (HTO) 1984 7.6 E+03 Lognormal, GSD = 13 N Reactor, PUREX

Tritium (HTO) 1985 1.1 E+03 Lognormal, GSD = 3.3 N Reactor, PUREX

Tritium (HTO) 1986–1987 6.1 E+02 Lognormal, GSD = 6.2 N Reactor, PUREX 

Tritium (HTO) 1988–1992 6.1 E+02 Lognormal, GSD = 6.2 PUREX only

Tritium (HTO) 1978–1985 4.8 E+04 Lognormal, GSD = 3 Ingestion of drinking 

water in the 400 Area

U-233 (type F) + 

contaminants

1965, 1966, 1970 16 Constant PUREX

U-233 (type S) 1965, 1966 Use Pu intakes from 

Attachment C

From Attachment C, 

50th percentiles with 

GSDs

234-5Z

Am-241 (type M) July 1, 1972-

present

Use Pu intakes (type 

M) from Attachment 

C

50th percentiles with 

GSDs

Cm-244 (type M) July 1, 1972-

1973

Use Pu intakes (type 

M) from Attachment 

C

95th percentiles, 

constant distribution

Cm-244 (type M) 1974-present Use Pu intakes (type 

M) from Attachment 

C

50th percentiles with 

GSDs

Np-237 (type M) July 1, 1972-

1983

Use Pu intakes (type 

M) from Attachment 

C

95th percentiles, 

constant distribution

Np-237 (type M) 1984-present Use Pu intakes (type 

M) from Attachment 

C

50th percentiles with 

GSDs

Po-210 (type F or 

M)

July 1, 1972-

present

Use Pu intakes (type 

M) from Attachment 

C

50th percentiles with 

GSDs

Thorium (type M 

or S)

July 1, 1972-

present

Use U intakes (type 

M or S) from 

Attachment C

95th percentiles, 

constant distribution

See Section 5.3.4 for 

progeny intakes

Tc-99 (type F or 

M)

July 1, 1972-

present

None None No known exposures 

during this period. 

D&D radiological 

workers

2002–present See discussion, last 

paragraph in this 

section. 

As presented in Section 5.1.2, 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 (72 FR 55214). 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. 

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 

as presented in Battelle-TBD-6000/PNWD-3738, Technical Basis Document: Site Profiles for Atomic 

Weapons Employers That Worked Uranium and Thorium Metals (Battelle 2006).

Attachment C intake rates exhibit a decreasing trend over time and may be assigned past 1988 when 

overestimates are appropriate for dose reconstructions. 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 64 of 163

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 2007a) and 131I and 

tritium intakes from Table 5-31.

When not operating, 1988 to the present

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 2007a) and tritium for 

PRTR from Table 5-31.

: 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.

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 (2007a). Note that this 

reactor continued to be used intermittently until 1994. Use tritium ingestion from drinking water from 

Table 5-31.

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 Super S), and uranium (type F from 

January 1, 1949 on) from Attachment C (except 147Pm) and ORAUT (2007a) but may ignore 

24Na from Attachment C for best estimates. May use 233U (and contaminants) intakes for 

PUREX 1965-70 from Table 5-31 in lieu of plutonium and typical uranium. Use tritium and 131I 

intakes from Table 5-31. No short-lived fission and activation products should be applied after 

1988 for best estimates.

B-Plant, 225-B, 1953 to the present

Plutonium Finishing Plants (various Z buildings and 233-S), September 1, 1946 to present: Use 

plutonium intakes from Attachment C (M, S, or Super S). If it is more favorable to the claimant:

: Use fission and activation products from Attachment C

(except 147Pm) and ORAUT (2007a – waste management option) but may ignore 24Na for best 

estimates. Use 65Zn ingestion-only intakes for 1953 to 1972. 

• For July 1, 1972 through 1976, intakes of 241Am may be assigned in lieu of plutonium.

• For 234-5Z, for 1958, 1965, and 1966, intakes of 233U with contaminants may be assigned in 

lieu of plutonium. 

• For 231-Z, 241Am may be assigned in lieu of plutonium (from July 1, 1972 on) and the same for 233U in 1957 and 1958.

221-U, 224-U (UO3 Plant), 1952 to 1993: Use uranium type F intakes from Attachment C. 

WR Vault, 1965-1979: If evidence of involvement with storing or shipping of thorium (primarily 

storage and transfer of liquid thorium nitrate solution), from July 1, 1972 on, use uranium type F or M 

from Attachment C. 

222-S Laboratory: Use fission and activation products (fuel dissolution) from Attachment C and 

ORAUT (2007a). Include 147Pm for 1964 through 1965. 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 65 of 163

108-B Tritium Extraction Facility, 1949 to 1956: Use tritium intakes from Table 5-31. Exposure to 

organically bound tritium 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 (2007a); 24Na may be ignored for best estimates. Use plutonium (M, S, 

Super S) and uranium (F) from Attachment C.

1962 to 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/PNWD-3738, Technical Basis Document: Site 

Profiles for Atomic Weapons Employers That Worked Uranium and Thorium Metals (Battelle 2006). 

: Use fission and activation products from Attachment C and ORAUT (2007a) 

(fuel dissolution), 24Na may be ignored for best estimates. Use 65Zn ingestion intakes only. 

Include 147Pm for 1964 through 1967.

209-E and 120 Critical Mass Laboratories, 1950 to 1986: Use plutonium intakes (M, S, or Super S) 

from Attachment C.

308 Plutonium Fuels Pilot Plant, 1960 to 1990: Use fission and activation products (fuel 

dissolution), plutonium (M, S, or Super S), and uranium (M or S), and 147Pm intakes from Attachment 

C and ORAUT (2007a). Use 131I intakes from Table 5-31; 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 (2007a). The fuel was well-aged before arriving, so 24Na and other shortlived radionuclides in ORAUT (2007a) may be ignored for best estimates and may use 65Zn ingestion 

intakes only. From July 1, 1972 on, include intakes of 241Am at plutonium coworker intake levels from 

Attachment C.

325 Building, 1953 to the present: Use fission and activation products (fuel dissolution), plutonium 

(M, S, or Super S), and 147Pm intakes from Attachment C and ORAUT (2007a); 24Na may be ignored 

for best estimates; however, 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 Super S) intakes from Attachment C and ORAUT (2007a). 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 Super S) intakes from Attachment C (except 147Pm) and ORAUT (2007a). Assign 

uranium intakes (types M or S from January 1, 1949 on) from Attachment C. 

3732 Building (Process Equipment Development Laboratory) and 3722 Building, July 1, 1972-

1980: Note: thorium work ceased in 1970 but decontamination of these facilities occurred shortly after 

shutdown. The time period has not been determined. Exposure prior to July 1, 1972 is covered by 

the SEC because it was determined to be infeasible to reconstruct doses, so no intakes are assigned. 

Assign thorium based on uranium coworker intakes from Attachment C (type, F, M, or S).
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 66 of 163

3708 Building: Assign intakes of 241Am or 244Cm based on plutonium coworker 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 2007a). Assign plutonium (M, S, or Super S) or americium 

intakes at 0.15 times the 137Cs fission product intake from Attachment C (to account for an alpha 

component in the waste). 24Na 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; however, 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 Super S) [122].

Rovers: Could have been exposed at many different facilities. Use the same intakes as for the fuel 

separations facilities, except do not include 233U.

Unknown facility: Use judgment based on any available information and ORAUT (2007a). 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. Mr. Bihl continues to serve as 

a Site Expert for this document because he possesses or is aware of information relevant for 

reconstructing radiation doses experienced by claimants who worked at the site. 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. Mr. Bihl’s Disclosure Statement is available at www.oraucoc.org. 

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. Mr. Scalsky continues to participate on this document team in the 

appropriate role of Subject Expert in compliance with the NIOSH Conflict or Bias policy.
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 67 of 163

[1] Bihl, Donald E. Battelle Memorial Institute. 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. Battelle Memorial Institute. 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. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on discussions with former in vivo counting program staff E. H. Palmer and G. A. 

Rieksts. The GOK region also shows on whole-body count forms from the 1960s and 1970s.

[4] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. 

The dates for HIE and ORE 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. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on personal experience and review of the bioassay database.

[6] Bihl, Donald E . Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on personal experience and review of the bioassay database.

[7] Bihl, Donald E. Battelle Memorial Institute. 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. Battelle Memorial Institute. 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. Battelle Memorial Institute. 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 Personnel Dosimetry Program Manager, K. Heid, was asked that 

direct question shortly after his retirement in the late 1980s.

[10] Bihl, Donald E. Battelle Memorial Institute. 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. Battelle Memorial Institute. 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 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 68 of 163

also verified when the former Personnel Dosimetry Program Manager, K. Heid, was asked that 

direct question shortly after his retirement, in the late 1980s.

[12] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. November 2006. 

This information was provided by J. A. MacLellan, the Bioassay Laboratory Contract Technical 

Administrator, and C. L. Antonio, Internal Dosimetrist 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. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on personal experience.

[14] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on personal experience and discussions with J. A. MacLellan, the Bioassay 

Laboratory Contract Technical Administrator.

[15] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on personal experience. In addition, a summary of bioassay procedures is 

contained in PNNL (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.

[16] Bihl, Donald E. Battelle Memorial Institute. 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. Battelle Memorial Institute. 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. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This information was extracted from various statements of work with the bioassay laboratory, 

including AEC (1964), ERDA (1975), DOE (1979), Simpson (1983), Author unknown (1985, 

1987), and Acker (1992).

[19] Bihl, Donald E. Battelle Memorial Institute. 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.

[20] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on discussions with PUREX engineers in the 1980s and from Sula, Carbaugh, 

and Bihl (1989).

[21] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

The composition of the plutonium used in the Nuclear Waste Vitrification Project was added to 

the Internal Dosimetry Program technical basis document 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.
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 69 of 163

[22] Bihl, Donald E. Battelle Memorial Institute. 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.

[23] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. November 2004. 

Information in the previous five sentences was based on a query of the REX database.

[24] Bihl, Donald E. Battelle Memorial Institute. 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.

[25] Bihl, Donald E. Battelle Memorial Institute. 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.

[26] Bihl, Donald E. Battelle Memorial Institute. 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.

[27] Bihl, Donald E. Battelle Memorial Institute. 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 5-1, the discussion in the plutonium section, and 

Attribution 12.

[28] Bihl, Donald E. Battelle Memorial Institute. 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.

[29] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on personal experience.

[30] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC (1964), ERDA (1975), DOE (1979), Simpson (1983), Author 

unknown (1985, 1987), and Acker (1992).

[31] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC (1964), ERDA (1975), DOE (1979), BPNL (1982), Simpson (1983), 

Author unknown (1985, 1987), Acker (1992), and PNNL (2003b, Appendix B).

[32] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This information was provided by J. A. MacLellan, the Bioassay Laboratory Contract Technical 

Administrator, and C. L. Antonio, Internal Dosimetrist in the Hanford Internal Dosimetry 

Program. It is also discussed in a letter to the Hanford Internal Dosimetry Program file 

(Antonio 2002).

[33] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This information was compiled from the various statements of work with the bioassay 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 70 of 163

laboratory, including AEC (1964), Lardy (1970), ERDA (1975), DOE (1979), BPNL (1982), 

Simpson (1983), Author unknown (1985, 1987), and Acker (1992).

[34] Bihl, Donald E. Battelle Memorial Institute. 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.

[35] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC (1964), Lardy (1970), ERDA (1975), DOE (1979), BPNL (1982), 

Simpson (1983), Author unknown (1985, 1987), and Adkins (1992).

[36] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. January 2009.

Based on a conversation with W. Baumgartner, who worked as a radiation monitor for a time 

at the 108-B tritium plant

[37] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. January 2009

Based on a conversation with W. Baumgartner, who worked as a radiation monitor for a time 

at the 108-B tritium plant.

[38] Bihl, Donald E. Battelle Memorial Institute. 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.

[39] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. November 2006. 

Because both the DOE Hanford RU document (DOE 2000) and the Hanford internal dosimetry 

manual (PNNL 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.

[40] Bihl, Donald E. Battelle Memorial Institute. 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).

[41] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on personal experience.

[42] Bihl, Donald E. Battelle Memorial Institute. 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.

[43] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

The rapid method was coded QUS for “quick uranium soluble.” Sula, Carbaugh, and Bihl 

(1989, Appendix C) refers to it as the “less sensitive method.” The kit code for the QUS was 7.
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 71 of 163

[44] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC (1964), Lardy (1970), ERDA (1975), DOE (1979), BPNL (1982), 

Simpson (1983), Author unknown (1985, 1987), and Adkins (1992).

[45] Bihl, Donald E. Battelle Memorial Institute. 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. Someone 

probably forgot to mark the special units on the uranium bioassay form.

[46] Bihl, Donald E. Battelle Memorial Institute. 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.

[47] Bihl, Donald E. Battelle Memorial Institute. 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.

[48] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This was explained to 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 awhile, and continued throughout the duration of the contract.

[49] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC (1964), Lardy (1970), ERDA (1975), DOE (1979), BPNL (1982), 

Simpson (1983), Author unknown (1985, 1987), and Adkins (1992).

[50] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

No use of strontium titanate at Hanford has been discovered.

[51] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC (1964), Lardy (1970), ERDA (1975), DOE (1979), BPNL (1982), 

Simpson (1983), Author unknown (1985, 1987), and Adkins (1992).

[52] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This information is based on query of the database and review of the data.

[53] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC (1964), Lardy (1970), ERDA (1975), DOE (1979), BPNL (1982), 

Simpson (1983), Author unknown (1985, 1987), and Adkins (1992).
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 72 of 163

[54] Bihl, Donald E. Battelle Memorial Institute. 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.

[55] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC (1964), Lardy (1970), ERDA (1975), DOE (1979), BPNL (1982), 

Simpson (1983), Author unknown (1985, 1987), and Adkins (1992).

[56] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This information was compiled from the various statements of work with the bioassay 

laboratory, including AEC (1964), Lardy (1970), ERDA (1975), DOE (1979), BPNL (1982), 

Simpson (1983), Author unknown (1985, 1987), and Adkins (1992).

[57] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2005. 

This is based on queries of the REX database.

[58] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2005. 

This is based on queries of the REX database.

[59] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on discussion with the project manager, Dr. D. Fisher, a PNNL employee.

[60] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on a casual discussion with Dr. M. Sikov, a retired PNNL radiobiologist, now 

deceased.

[61] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

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.

[62] Bihl, Donald E. Battelle Memorial Institute. 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 (Author unknown 1964). Some 

of the detection levels were also shown, with calculations, in Palmer’s 1960 Laboratory Record 

Notebook. Based on the calculations, 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).

[63] Bihl, Donald E. Battelle Memorial Institute. 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.

[64] Bihl, Donald E. Battelle Memorial Institute. 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.

[65] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

The reporting of these radionuclides is evident from a query on the database. It was also 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 73 of 163

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 Bihl, there were some counts for which one or more of 

the four regularly reported radionuclides were missing. So as a general rule they were nearly 

always reported, but clearly there are some exceptions in the database.

[66] Bihl, Donald E. Battelle Memorial Institute. 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.

[67] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on personal discussions with former in vivo counting program staff, H. E. Palmer 

and G. A. Rieksts. Palmer was the in vivo counting program manager for many years, and 

Rieksts was a scientist in the program for many years. Both are retired. The GOK region is

also shown on the early whole-body counting results forms.

[68] Bihl, Donald E. Battelle Memorial Institute. 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

[69] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on personal discussions with H. E. Palmer and G. A. Rieksts.

[70] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on personal experience and discussion with G. Rieksts. In addition, see for 

instance the 1992 In Vivo Bioassay Statement of Work (Lynch et al. 1992, p. 3.6).

[71] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on discussion with program managers H. E. Palmer and T. P. Lynch. In 

addition, it is implied in most of the annual reports, such as Lyon et al. (1988) and others.

[72] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on personal experience and discussion with H. E. Palmer.

[73] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2005. 

This is based on queries on the REX database for 131I results.

[74] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2005. 

This is based on discussion with T. P. Lynch and G. A. Rieksts.

[75] Bihl, Donald E. Battelle Memorial Institute. 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).

[76] Bihl, Donald E. Battelle Memorial Institute. 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 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 74 of 163

decision level; therefore, the recommendation was made to assume that the MDA was twice 

the reported detection level.

[77] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. November 2006. 

This is based on discussion with T. P. Lynch; however, this is an educated guess by Lynch 

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 the dose reconstructors to 

revise the 232Th results based on disequilibrium factors in the text.

[78] Traub, Richard. Battelle Memorial Institute. Principal Health Physicist. November 2006. 

This is based on calculations by R. Traub using MatlabTM. The calculations incorporated the 

latest ICRP biokinetic models with independent systemic biokinetics for the major progeny.

[79] Bihl, Donald E. Battelle Memorial Institute. 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, then 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.

[80] Traub, Richard. Battelle Memorial Institute. Principal Health Physicist. December 2006. 

These were calculated by R. Traub using MatlabTM. The calculations incorporated the latest 

ICRP biokinetic models with independent systemic biokinetics for the major progeny.

[81] Traub, Richard. Battelle Memorial Institute. Principal Health Physicist. December 2006. 

These were calculated by R. Traub using MatlabTM. The calculations incorporated the latest 

ICRP biokinetic models with independent systemic biokinetics for the major progeny.

[82] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. December 2006. 

Because of the very high activities of fission products that were handled in the PUREX 

canyon, the ventilation system was built to ensure exhaust of contaminants through the 

canyon filtration and exhaust system. The air flow rate through the canyon exhaust stack was 

nominally 100,000 ft3

/min. Fuel with high thorium and 233U content was dissolved in the 

canyon in the same way as normal fuel and thoron was exhausted out the canyon stack.

[83] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on personal experience and also evident from a query of the REX database.

[84] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on personal experience and discussion with T. Lynch, in vivo monitoring 

program manager. The correction for shine from the liver is also mentioned in BPNL (1990).

[85] Bihl, Donald E. Battelle Memorial Institute. 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.
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 75 of 163

[86] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on discussion with T. Lynch, in vivo monitoring program manager. 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.

[87] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is evident from a review of various whole-body count results forms.

[88] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

These changes were evident from review of the REX database.

[89] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

Implementation of NEXEC is discussed in Lyon et al. (1996). PNL (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.

[90] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on personal experience and discussion with T. Lynch.

[91] Bihl, Donald E. Battelle Memorial Institute. 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.

[92] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. May 2006. 

The FFTF was sodium cooled, so neutron activation of the sodium was a potential source of 

contamination (Section 2.1.9).

[93] Bihl, Donald E. Battelle Memorial Institute. 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.

[94] Napier, Bruce A. Battelle Memorial Institute. Principal Health Physicist. December 2008.

B. A. 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 

(http://twins.pnl.gov/twins3/twins.htm). 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 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 gammaemitting radionuclides.
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 76 of 163

[95] Bihl, Donald E. Battelle Memorial Institute. 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.

[96] Bihl, Donald E. Battelle Memorial Institute. 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.

[97] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on discussion with the in vivo monitoring program manager T. Lynch.

[98] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on discussion with the in vivo monitoring program manager T. Lynch.

[99] Bihl, Donald. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on discussion with the in vivo monitoring program manager T. Lynch.

[100] Bihl, Donald E. Battelle Memorial Institute. 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.

[101] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. October 2003. 

This is based on discussion with the in vivo monitoring program managers T. P. Lynch and 

G. A. Rieksts.

[102] Bihl, Donald E. Battelle Memorial Institute. 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 such that 

they were no longer significant contributors to contamination.

[103] Bihl, Donald E. Battelle Memorial Institute. 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 

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.
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 77 of 163

[104] Bihl, Donald E. Battelle Memorial Institute. 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.

[105] Bihl, Donald E. Battelle Memorial Institute. 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.

[106] Bihl, Donald E. Battelle Memorial Institute. 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.

[107] Bihl, Donald E. Battelle Memorial Institute. 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.

[108] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. May 2006. 

This is based on query of the REX database.

[109] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. May 2006. 

This is based on query of the REX database.

[110] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. May 2006. 

This was calculated using IMBA.

[111] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. May 2006. 

There were considerable differences in potential exposure to tritium between the 108B 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 108B. However, lacking air sample data, the order of magnitude ratio was 

considered reasonable and favorable to claimants.

[112] Bihl, Donald E. Battelle Memorial Institute. 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 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 Sparks.

[113] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. May 2006. 

There were considerable differences in potential exposure to tritium between the 108B 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 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 78 of 163

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 108B. However, lacking air sample data, the order of magnitude ratio was 

considered reasonable and favorable to claimants.

[114] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. November 2006. 

This was calculated using IMBA.

[115] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist; Joseph Lochamy. 

Scientist. XCEL Engineering, Inc.; and Richard Sparks. CDE Dosimetry Services, Inc. 

Principal Health Physicist. December 2006. 

The tritium dose data is from a query on REX by 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 Sparks.

[116] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. December 2006. 

These were calculated using IMBA.

[117] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. December 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.

[118] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. June 2005. 

These were calculated using IMBA. Ingestion of 40,000 pCi/d produces approximately 

1 mrem/yr to all organs.

[119] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. June 2006. 

There were not enough samples for a formal coworker analysis (n=11). The units were not 

known exactly; based on the magnitude of the results, microcuries per liter was assumed (see 

Section 5.2.5).

[120] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. June 2006. 

This was calculated using IMBA.

[121] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. June 2006. 

This is a summary of information developed in the text.

[122] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. June 2006. 

This is based on discussion with the D&D contractor’s dosimetry expert E. W. Carlson. 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 (PNNL 2003b).

[123] Bihl, Donald E. Battelle Memorial Institute. 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. 04 Effective Date: 10/20/2010 Page 79 of 163

[124] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. 

This is from a query on the REX database by the REX custodian.

[125] Arno, Matthew. Foxfire Scientific. Dose Reconstructor. June 11, 2007.

The periods for how data was recorded are taken directly from examination of the HERB 

database.

[126] Bihl, Donald E. Battelle Memorial Institute. 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 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 coworker analysis were excluded.

[127] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. July 2005.

The promethium intakes were identified by the patterns of excretion that were associated with 

each individual and followed by review of each individual’s radiological exposure file, to which 

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 coworker analysis were 

excluded.

[128] Arno, Matthew. Foxfire Scientific. 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.

[129] Arno, Matthew. Foxfire Scientific. Dose Reconstructor. June 2007. 

Ten-year aged fuel-grade plutonium is the default assumption per the main text of this 

document. However, before 1955, it is not possible to have 10-year aged plutonium. 

Therefore, the most favorable to claimant credible mixture is used.

[130] Arno, Matthew. Foxfire Scientific. 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.

[131] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. July 2005.

Plot created by Bihl from data in Heeb (1994, Appendix A).

[132] Bihl, Donald E. Battelle Memorial Institute. Principal Health Physicist. March 2009.

The Hanford Site Description (ORAUT 2010c) 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 1-F 

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-28 for the period of major 

production from 1966 to 69. Considering that the February 1963 bioassays resulted from a 

major contamination spread, and that work conditions were similar in the two periods, it was 
Document No. ORAUT-TKBS-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 80 of 163

concluded that the 1966 to 1969 coworker 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. 04 Effective Date: 10/20/2010 Page 81 of 163

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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

Hanford term, circa 1940s and 1950s, meaning 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 (α)

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

The driving off of all moisture in a sample; accomplished by heating in an oven or by dissolving 

the sample in a liquid (often nitric acid), then evaporating to dryness. The latter technique is 

called wet ashing. 

atom

Smallest amount of an element that cannot be chemically divided or broken up. It consists of 

a central core (the nucleus) surrounded by one or more electrons. See element and nucleus. 

beta particle (β)

See beta radiation. 

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

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, the 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.
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dose

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of roentgens, rads, reps, 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. 

electromagnetic radiation

Energy that results from the acceleration of electric charge and the associated electric and 

magnetic fields. The energy can be regarded as waves propagated through space or as a 

stream of photons. The electromagnetic radiation spectrum ranges from short wavelengths 

(X- and gamma rays) to intermediate wavelengths (ultraviolet, visible, and infrared light) to 

long wavelengths (radar and radio). Only the higher energy (shorter wavelength) forms of 

electromagnetic radiation can ionize matter. Lower energy electromagnetic radiation, such as 

visible and infrared light, radar, and radio waves, does not have enough energy to ionize 

matter.

electron

Basic atomic particle with negative charge and a mass 1/1,837 that of a proton. Electrons 

surround the positively charged nucleus of the atom. See element. 

element

One of the known chemical substances in which the atoms have the same number of protons. 

Elements cannot be broken down further without changing their chemical properties. 

Chemical symbols for the elements consist of either a single letter or a combination of letters, 

some of which descend from the Latin names [e.g., Au from aurum (gold), Fe from ferrum

(iron)]. This glossary indicates elements by their names. Specific isotopes appear as their 

standard chemical symbols with the number of protons and neutrons in the nucleus. For 

example, the isotope of uranium that contains 92 protons and 143 neutrons can appear as 

235U, U-235, or uranium-235. See radioactive isotope. 

Energy Employees Occupational Illness Compensation Program Act of 2000, as amended 

(EEOICPA; 42 U.S.C. § 7384 et seq.) 

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.
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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 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

The measurements of radioactive material in the human body utilizing instrumentation that 

detects radiation emitted from the radioactive material in the body.

ionization

Process by which a neutral atom or molecule loses or gains electrons and therefore acquires a 

net electrical charge. See ionizing radiation. 

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. See 

alpha radiation, beta radiation, gamma radiation, neutron radiation, and X-ray radiation. 

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties. See element. 

nucleus

Central core of an atom, which consists of positively charged protons and, with the exception 

of ordinary hydrogen, electrically neutral neutrons. The number of protons (atomic number) 

uniquely defines a chemical element, and the number of protons and neutrons is the mass 

number of a nuclide. The plural is nuclei.
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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. 

National Institute for Occupational Safety and Health (NIOSH)

U.S. agency responsible for dose reconstruction under the Energy Employees Occupational 

Illness Compensation Program. 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.

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.

operating area

Designation of Hanford major operational work areas among the respective fuel fabrication 

(e.g., 300 Area), reactor operations (e.g., 100B, 100C, 100D, 100DR, 100F, 100H, 100KE, 

100KW, and 100N), chemical separations (e.g., U-, T-, B-, UO3, REDOX, and PUREX Plants), 

plutonium finishing (Z-Plant), research and development (e.g. 300 and 3000 Areas), and 

transportation, communication, and general site support (e.g., 600, 700, and 1100 Areas).

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 (POC, PC)

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Act, 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. 

radioactive isotope

Natural or synthetic form of an atom that emits radioactivity when it decays. See isotope. 

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 radioactive and 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

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 83 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 Xor gamma rays. Replaced by the gray in the International System of Units; 1 rep is 

approximately equal to 8.38 milligray. The word derives from roentgen equivalent physical.

simulated 24-hour urine sample

Collection of all urine samples beginning with the void before retiring for the evening and 

ending with the first void after rising the next morning, for two consecutive nights, to simulate a 

24-hour urine sample.

Special Exposure Cohort (SEC) [42 U.S.C. § 7384l(14)]

... “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.
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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 

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 isotopic mixture with nominally 4% to 6% 240Pu.

whole-body counter (WBC)

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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Various health physics textbooks and histories document the history of the tolerance dose. The 

description below was extracted from Jacob Shapiro’s textbook (Shapiro 1990) and from 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 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) states 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/cc 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, the following are examples of the “tolerance limits for prolonged exposure” 

at Hanford in 1945 (Cantril 1945):

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/cc

Radioactive I-131 in atmosphere 1.0 × 10-13 Ci/cc

Mixed fission products in drinking water 1.2 × 10-9 Ci/cc

Productc concentration in drinking water 10-5 μg/cc

Product concentration in atmosphere 5 × 10-10 μg/cc

Product deposition in body 0.5-μg total accumulation

Uranium dust in atmosphere 1.5 × 10-4 μg/cc

a. The rep (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 [123].

c. “Product” was a euphemism for plutonium.

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 

ATTACHMENT A

TOLERANCE DOSE AND TOLERANCE VALUES AT HANFORD

Page 1 of 2 
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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/liter).

Mixed fission products: 0.1 μc/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 μc/L [123] 

Other air contamination limits included:

• Tritium: Gas 0.1 μc/L, vapor 0.01 μc/L

• Carbon: 0.02 μc/L

• Argon: 1.6 × 10-3 μc/L 

• Iodine – 1.5 × 10-6 μc/L

• Xenon: 0.01 μc/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 a single 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 a single 

8-hour exposure to be 1.2 × 10 11 Ci/cm3

. 

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TOLERANCE DOSE AND TOLERANCE VALUES AT HANFORD

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LIST OF TABLES 

TABLE TITLE 

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

PAGE

100

B-2 Sample type codes............................................................................................................... 103

B-3 Bioassay reason codes ........................................................................................................ 104

B-4 Excreta sample kit codes ..................................................................................................... 105

B-5 In vivo body location codes .................................................................................................. 106

B-6 Excreta unit codes................................................................................................................ 106

B-7 Excreta processing codes .................................................................................................... 106

B-8 Excreta laboratory codes...................................................................................................... 107

B-9 Excreta no-sample codes..................................................................................................... 107

B-10 In vivo invalid result codes ................................................................................................... 107

B-11 INTERTRAC mode-of-intake codes ..................................................................................... 108

B-12 INTERTRAC evaluation reason codes ................................................................................. 108

B-13 INTERTRAC source-of-intake codes.................................................................................... 108

B-14 INTERTRAC miscellaneous codes....................................................................................... 108

B-15 Special whole-body count resolution codes.......................................................................... 109

This attachment is adapted from PNNL (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 [124]. 

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

ACS Actinium-227, thorium-227 Scheduling code

AC227 Actinium-227

AC228 Actinium-228

AG110 Silver-110

AM241 Americium-241

AM242 Americium-242

AM243 Americium-243

BA140 Barium-140

BETA Beta

BI213 Bismuth-213

BI214 Bismuth-214

BK249 Berkelium-249

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Code Description Comment

BR 82 Bromine-82

C 14 Carbon-14

CE141 Cerium-141

CE143 Cerium-143

CE144 Cerium-144

CF249 Californium-249

CM242 Curium-242

CM244 Curium-244

CO 58 Cobalt-58

CO 60 Cobalt-60

CR 51 Chromium-51

CS134 Cesium-134

CS137 Cesium-137

EU152 Europium-152

EU154 Europium-154

EU155 Europium-155

EU156 Europium-156

EV155 ? Probably a typographical error for Eu-155 that got left in the 

database 

EV156 ? Probably a typographical error for Eu-156 that got left in the 

database 

FE 59 Iron-59

FP Fission products

GA Gross alpha

GB Gross beta

GELI Gamma-GeLi detector Excreta scheduling code for a gamma scan with a germanium 

detector

GOK See text

GS Gamma NaI detector Excreta scheduling code for a gamma scan with a NaI detector

H 3 Tritium

I 125 Iodine-125

I 129 Iodine-129

I 131 Iodine-131

I 133 Iodine-133

IAM Isotopic americium Excreta scheduling code for americium separation and alpha 

spectrometry

ICA ? 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 separate anal. by beta counting

IPUBA Pu isotopic, Pu-241, Am-241 Excreta scheduling code

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Code Description Comment

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 & Am-241 Excreta scheduling code

IU U isotopic Excreta scheduling code

IUPU Isotopic plutonium/NU Excreta scheduling code

K 40 Potassium

LA140 Lanthanum-140

LEPD Low-energy photon detector Excreta scheduling code for low-energy photon scan

MFP Mixed fission products

MN 54 Manganese-54

MO 99 Molybdenum-99

NA 22 Sodium-22

NA 24 Sodium-24

NAI Gamma NaI detector Excreta scheduling code

NB 95 Niobium-95

NP237 Neptunium-237

NP239 Neptunium-239

PB210 Lead-210

PB212 Lead-212

PM147 Promethium-147

PO210 Polonium-210

PR144 Praseodymium-144

PU Plutonium alpha Total alpha from Pu isotopes after separation

PUMIX Plutonium alpha Total alpha from Pu isotopes and Am-241

PU238 Plutonium-238

PU239 Plutonium-239 When pertaining to excreta samples, it’s actually Pu-239+240

PU240 Plutonium-240

PU241 Plutonium-241

PU242 Plutonium-242

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

RA225 Radium-225

RA226 Radium-226

RA228 Radium-228

RH106 Rhodium-106

RND Radon daughters

RU103 Ruthenium-103

RU106 Ruthenium-106

S 35 Sulfur-35

SB124 Antimony-124

SB125 Antimony-125

SCP Sequential Sr-total Ce Pm Excreta scheduling code

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Code Description Comment

SM153 Samarium-153

SR Strontium Total radiostrontium by beta counting

SR 89 Strontium-89

SR 90 Strontium-90 When pertaining to excreta samples, Sr-90 by yttrium ingrowth

TAC Total actinides

TC 99 Technetium-99

TH227 Thorium-227

TH228 Thorium-228

TH230 Thorium-230

TH232 Thorium-232

TH234 Thorium-234

TL208 Thallium-208

U Elemental uranium

URAN Elemental uranium

U DEP Depleted uranium

U NAT NU

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

U 236 Uranium-236

U 238 Uranium-238

UMIX Uranium mix Total uranium, used for intakes not bioassay

UMS U-235 U-236 U-238 U-234

US U

XX 0 Isotope will have no result

ZN 65 Zinc-65

ZR 95 Zirconium-95

Table B-2. Sample type codes. 

Code Type of sample

B Blood

F Feces

S Sputum

T Tissue

U Urine

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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 

follow-up 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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Table B-4. Excreta sample kit codesa. 

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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CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS

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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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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 PNNL 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 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.

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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

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 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

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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

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TABLE OF CONTENTS

SECTION TITLE

C.1 Purpose ...............................................................................................................................

PAGE

114

C.2 Overview.............................................................................................................................. 114

C.3 Data ..................................................................................................................................... 117

C.3.1 Selected Bioassay Data............................................................................................117

C.3.1.1 Plutonium Urinalysis................................................................................... 117

C.3.1.2 Uranium Urinalysis ..................................................................................... 119

C.3.1.3 Strontium Urinalysis ................................................................................... 121

C.3.1.4 Promethium Urinalysis ............................................................................... 122

C.3.1.5 Zinc-65 in Whole-Body Counts................................................................... 123

C.3.1.6 Sodium-24 in Whole-Body Counts ............................................................. 124

C.3.1.7 Cesium-137 in Whole-Body Counts ........................................................... 125

C.3.2 Analysis....................................................................................................................126

C.3.3 Tritium Dose Evaluation............................................................................................126

C.4 Intake Modeling.................................................................................................................... 129

C.4.1 Assumptions.............................................................................................................129

C.4.2 Bioassay Fitting ........................................................................................................130

C.4.3 Radionuclides and Material Types............................................................................ 130

C.4.3.1 Plutonium................................................................................................... 130

C.4.3.1.1 Type M ..................................................................................... 130

C.4.3.1.2 Type S...................................................................................... 131

C.4.3.2 Uranium ..................................................................................................... 131

C.4.3.2.1 Type F ...................................................................................... 131

C.4.3.2.2 Type M ..................................................................................... 132

C.4.3.2.3 Type S...................................................................................... 132

C.4.3.2.4 Strontium-90 ............................................................................. 133

C.4.3.3 Promethium-147 ........................................................................................ 133

C.4.3.3.1 Type M ..................................................................................... 133

C.4.3.3.2 Type S...................................................................................... 133

C.4.3.4 Zinc-65....................................................................................................... 133

C.4.3.4.1 Inhalation Intakes...................................................................... 134

C.4.3.4.2 Ingestion Intakes....................................................................... 134

C.4.3.5 Sodium-24.................................................................................................. 134

C.4.3.5.1 Inhalation Intakes...................................................................... 134

C.4.3.5.2 Ingestion Intakes....................................................................... 134

C.4.3.6 Cesium-137................................................................................................ 135

C.5 Assignment of Intakes and Doses ........................................................................................ 135

C.5.1 Plutonium .................................................................................................................137

C.5.2 Uranium....................................................................................................................137

C.5.3 Strontium-90.............................................................................................................137

C.5.4 Promethium-147.......................................................................................................139

C.5.5 Zinc-65 .....................................................................................................................139

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C.5.6 Sodium-24................................................................................................................140

C.5.7 Cesium-137..............................................................................................................141

C.5.8 Additional Radionuclides ..........................................................................................141

LIST OF TABLES 

TABLE TITLE

C-1 Time intervals for statistical analysis of plutonium urine samples .........................................

PAGE

120

C-2 Time intervals for statistical analysis of uranium urine samples............................................ 121

C-3 Time intervals for statistical analysis of 90Sr urine samples .................................................. 122

C-4 Time intervals for statistical analysis of 147Pm urine samples ............................................... 123

C-5 Time intervals for statistical analysis of 65Zn whole-body counts .......................................... 124

C-6 Time intervals for statistical analysis of 24Na whole-body counts .......................................... 125

C-7 Time intervals for statistical analysis of 137Cs whole-body counts......................................... 125

C-8 Plutonium mixture usage...................................................................................................... 129

C-9 Type M 239+240Pu intake periods and rates ............................................................................ 131

C-10 Type S 239+240Pu intake period and rates .............................................................................. 131

C-11 Type F uranium intake periods and rates ............................................................................. 132

C-12 Type M uranium intake periods and rates ............................................................................ 132

C-13 Type S uranium intake periods and rates ............................................................................. 132

C-14 Type F 90Sr intake periods and rates.................................................................................... 133

C-15 Type M 147Pm intake periods and rates ................................................................................ 133

C-16 Type S 147Pm intake periods and rates................................................................................. 133

C-17 Type S zinc-65 inhalation intake periods and rates .............................................................. 134

C-18 Zinc-65 ingestion intake periods and rates ........................................................................... 134

C-19 Type F 24Na inhalation intake periods and rates................................................................... 135

C-20 Sodium-24 ingestion intake periods and rates...................................................................... 135

C-21 Type F 137Cs intake periods and rates.................................................................................. 135

C-22 Type M plutonium-239 intake periods and rates................................................................... 137

C-23 Type S plutonium-239 intake period and rates ..................................................................... 137

C-24 Uranium Type F intake periods and rates............................................................................. 138

C-25 Uranium Type M intake periods and rates............................................................................ 138

C-26 Uranium Type S intake periods and rates ............................................................................ 138

C-27 Type F 90Sr intake periods and rates.................................................................................... 139

C-28 Promethium-147 intake periods and rates............................................................................ 139

C-29 Zinc-65 intake periods and rates .......................................................................................... 140

C-30 Sodium-24 intake periods and rates..................................................................................... 140

C-31 Type F cesium-137 intake periods and rates........................................................................ 141

C-32 Statistical summary of 239Pu 24-hour urinary excretion rates, 1946 to 1988 ......................... 142

C-33 Statistical summary of uranium 24-hour urinary excretion rates, 1948 to 1988..................... 148

C-34 Statistical summary of 90Sr 24-hour urinary excretion rates, 1966 to 1988 ........................... 152

C-35 Statistical summary of 147Pm 24-hour urinary excretion rates, 1967 to 1977......................... 153

C-36 Statistical summary of 65Zn measured in whole-body counts, 1960 to 1984 ......................... 156

C-37 Statistical summary of 24Na measured in whole-body counts, 1960 to 1984......................... 160

C-38 Statistical summary of 137Cs measured in whole-body counts, 1960 to 1988........................ 162

C-39 Hanford monthly reports....................................................................................................... 163

C-40 Tritium hard-copy bioassay data .......................................................................................... 163

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INTERNAL DOSIMETRY COWORKER DATA FOR THE HANFORD SITE

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LIST OF FIGURES

FIGURE TITLE

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

PAGE

115

C-2 Uranium bioassay data in REX and from Hanford monthly reports....................................... 116

C-3 Tritium bioassay data........................................................................................................... 127

C-4 Individual tritium bioassay or dose data by year................................................................... 128

C-5 Tritium dose from hard-copy records and the main text........................................................ 128

C-6 Production rates in tons of uranium times fuel burnup, 1944 to 1972 ................................... 136

C-7 50th-percentile plutonium urinalysis data for intakes of Type M material, 1946 to 1948 ....... 143

C-8 50th-percentile plutonium urinalysis data for intakes of Type M material, 1949 to 1952 ....... 144

C-9 50th-percentile plutonium urinalysis data for intakes of Type M material, 1953 to 1981 ....... 144

C-10 50th-percentile plutonium urinalysis data for intakes of Type M material, 1982 to 1988 ....... 144

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

based on four independent intakes, compared to bioassay data .......................................... 145

C-12 50th-percentile plutonium urinalysis data for intakes of Type S material, 1946 to 1948........ 145

C-13 50th-percentile plutonium urinalysis data for intakes of Type S material, 1949 to 1952........ 146

C-14 50th-percentile plutonium urinalysis data for intakes of Type S material, 1953 to 1981........ 146

C-15 50th-percentile plutonium urinalysis data for intakes of Type S material, 1982 to 1988........ 146

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

based on four independent intakes, compared to bioassay data .......................................... 147

C-17 50th-percentile uranium urinalysis data for intakes of Type F material, 1948 to 1988........... 149

C-18 50th-percentile uranium urinalysis data for intakes of Type M material, 1948 to 1988.......... 149

C-19 50th-percentile uranium urinalysis data for intakes of Type S material, 1948 to 1952........... 149

C-20 50th-percentile uranium urinalysis data for intakes of Type S material, 1953 to 1956........... 150

C-21 50th-percentile uranium urinalysis data for intakes of Type S material, 1957 to 1961........... 150

C-22 50th-percentile uranium urinalysis data for intakes of Type S material, 1962 to 1974........... 150

C-23 50th-percentile uranium urinalysis data for intakes of Type S material, 1975 to 1983........... 151

C-24 50th-percentile uranium urinalysis data for intakes of Type S material, 1984 to 1988........... 151

C-25 Predicted 50th-percentile urinary excretion of Type S uranium from 1946 to 1988 

based on six independent intakes, compared to bioassay data............................................ 151

C-26 50th-percentile strontium urinalysis data for intakes of Type F material, 1965 to 1988......... 152

C-27 50th-percentile promethium urinalysis data for intakes of Type M material, 1966 to 

1969..................................................................................................................................... 153

C-28 50th-percentile promethium urinalysis data for intakes of Type M material, 1970 to 

1979..................................................................................................................................... 153

C-29 Predicted 50th-percentile urinary excretion of Type M promethium from 1966 to 1979 

based on two independent intakes, compared to bioassay data........................................... 154

C-30 50th-percentile promethium urinalysis data for intakes of Type S material, 1966 to 

1969..................................................................................................................................... 154

C-31 50th-percentile promethium urinalysis data for intakes of Type S material, 1970 to 

1979..................................................................................................................................... 154

C-32 Predicted 50th-percentile urinary excretion of Type S promethium from 1966 to 1979 

based on two independent intakes, compared to bioassay data........................................... 155

C-33 50th-percentile zinc whole-body counting data for inhalation intakes of Type S 

material, 1960 to 1963 ......................................................................................................... 157

C-34 50th-percentile zinc whole-body counting data for inhalation intakes of Type S 

material, 1964 to 1967 ......................................................................................................... 157

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C-35 50th-percentile zinc whole-body counting data for inhalation intakes of Type S 

material, 1968 to 1970 ......................................................................................................... 157

C-36 50th-percentile zinc whole-body counting data for inhalation intakes of Type S 

material, 1971 to 1972 ......................................................................................................... 158

C-37 50th-percentile zinc whole-body counting data for inhalation intakes of Type S 

material, 1973 to 1977 ......................................................................................................... 158

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

material, 1978 to 1984 ......................................................................................................... 158

C-39 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....................................................................................................................... 159

C-41 50th-percentile sodium whole-body counting data for inhalation intakes of Type F 

material, 1960 to 1984 ......................................................................................................... 161

C-42 50th-percentile sodium whole-body counting data for ingestion intakes, 1960 to 1984......... 161

C-43 50th-percentile 137Cs whole-body counting data for inhalation intakes of Type F 

material, 1960 to 1988 ......................................................................................................... 163

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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 

coworker data from the Hanford site (including PNNL) for the purpose of estimating unmonitored 

exposures. Many of the periods that are addressed by this coworker analysis include those that are 

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

ORAUT-OTIB-0019, Analysis of Coworker Bioassay Data for Internal Dose Assignment (ORAUT 

2005), describes the general process for analyzing bioassay data for the assignment of doses based 

on coworker results.

Bioassay results for Hanford were obtained as a copy of the REX database sent to the NIOSH 

Health-Related Energy Research Branch (HERB) before the start of the dose reconstruction project. 

The REX data necessary to generate the bioassay statistics for the coworker 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-39 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 for exposure at Hanford 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 coworker study.

The urinalysis data that were chosen for this coworker study because of general applicability and 

number of measurements were from plutonium, uranium, radiostrontium, and promethium analyses, 

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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

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

although the 147Pm intakes have limited applicability (Sections C.3.1.4 and C.5.1.4). 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.” 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 (Section C.5).

With the exception of naturally-occurring 40K, only three gamma-emitting radionuclides that were 

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. 

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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

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

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 coworker analysis

of penetrating external dose and was therefore not duplicated in this coworker analysis. The 

completeness of those records was subsequently evaluated as discussed in Section C.3.3. [The 

coworker 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 (2005). The resultant values were input to the IMBA 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.

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C.3 DATA

C.3.1

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.

Selected Bioassay Data

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 

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 (HPS 1996). The analysis established time intervals for which 

single values for the 50th- and 84th-percentile values were calculated. These intervals were 

determined such that 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, and there was therefore 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 2005) for values that were 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 241

Am. Since the fourth quarter of 1983, the 

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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, which 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 

volumes of 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 follow-up 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 follow-up 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:

• If there were a large number of samples from a single worker, usually bunched after a specific 

time, these were removed. 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.

• If a record of an acute intake was judged to be unrepresentative, that record was removed. 

Any intake that was treated by diethylenetriaminepentaacetic 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 there were many 

sample results that were recorded exactly at the recording level but 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.

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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 

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.

The following were set as the top of the linear distribution:

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–Sep. 9, 1981 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, therefore, 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 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-1 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.

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 

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Table C-1. 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–Jun. 1981 Quarter years

3Q 1981 Jul. 1–Sep. 9

4Q 1981 Sep. 10–Dec. 31

1982–1988 Quarter years

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 

present. The latter were converted to micrograms per liter using the sample volume that was 

recorded as part of the information in the database [125]. Elemental uranium samples were not 

usually 24-hour samples; a mix of sample collection periods was used from 1947 to 1988 with 

overnight sampling being 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) [126].

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.

Reporting levels were: 

1948–1974 4.0 μg/L 

1975–1881 0.4 μg/L

1982 to 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 there were many results 

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.

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• 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-2 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). There 

were only 65 samples collected in 1973; the results were comparable to neighboring years, so the 

1973 statistical parameters were included.

Table C-2. 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 the 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 follow-up samples to known large acute intakes; one sample result was 

removed because it was a recount. There were some obvious duplicates that were removed.

Reporting levels were: 

1965–1969 1.67E-5 μCi//L

1970–1974 1.00E-6 μCi/L

1975–Mar. 1979 2 dpm/sample

Apr. 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.

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• 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-3 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. Whole-body counting was the primary bioassay 

method for fission products; the exception was for the workers at the WESF 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.

Table C-3. 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

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 (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 probably 

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.

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Most routine samples were simulated 24-hour samples, so samples with volumes smaller than 

400 mL were removed from the data before statistical analysis. There were a large number of 

duplicate entries in the database from 1966 to 1968, and these were 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 [127].

Reporting levels were:

1966–Mar. 25, 1970 1.67E-5 μCi/L

Mar. 31, 1970–1974 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-4 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-4. Time intervals for statistical analysis of 147Pm urine samples. 

Calendar year Analysis interval

1966, 1967 As one group

1968, 1969, 1970 through March 25 As one group

Mar. 31–Dec. 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 

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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.

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, with only one value 

that fell between zero and 0.75 nCi. Blanks and zeros were distributed using the linear distribution.

Table C-5 lists the time intervals for separate statistical analyses.

Table C-5. 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; after 1984 there were very few recorded results. 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 the 65Zn section, 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 

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numbers from the distribution. For 1975 to 1984, a value of 0.31 nCi was distinctly being used as the 

reporting level with no measurements recorded lower than 0.31 nCi. Blanks and zeros were 

distributed using the distribution with 0.3 nCi as the upper range.

Table C-6 lists the time intervals for separate statistical analyses.

Table C-6. 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

C.3.1.7 Cesium-137 in Whole-Body Counts

Cesium-137 was recorded routinely 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. 

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 cesium-137 data 

except for the period of 1984 through 1988.

Table C-7 lists the time intervals for separate statistical analyses.

Table C-7. 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).

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C.3.2

For each of the seven radionuclides, a lognormal distribution for the data in the intervals in Tables C-1 

to C-7 was assumed. The 50th- and 84th-percentile values were calculated using the method in 

ORAUT (2005). Tables C-33 to C-39 show the statistical analysis results for plutonium, uranium, 90Sr, 

147Pm, 65Zn, 24Na, and 137Cs.

Analysis

C.3.3

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 

because assumptions for this project are not necessarily consistent with those of a radiation 

protection program. 

Tritium Dose Evaluation

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-40 contains a list of the SRDB reference numbers for the 

hard-copy tritium bioassay data. Table C-39 contains a list of the SRDB reference numbers for the 

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 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 timeframe 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 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.

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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

Figure C-3. Tritium bioassay data. 

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). Relatively good 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.

Using an updated version of a tool for calculation of coworker doses from tritium bioassay for the 

Mound and LANL coworker 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 in all but 1 

year than the values calculated from the hard-copy records. The values in Table 5-30 are already 

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 coworker doses.

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Individuals by Year

0

50

100

150

200

250

300

350

400

1945 1950 1955 1960 1965 1970

Year

# individuals

REX Hardcopy Matched

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

Coworker Tritium Dose

0

20

40

60

80

100

120

140

160

180

200

CY 1949CY 1951CY 1953CY 1955CY 1957CY 1959CY 1961CY 1963CY 1965CY 1967CY 1969

Year

mrem

Hardcopy records TBD Values

Figure C-5. Tritium dose from hard-copy records and the main text (Section 5.6).

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C.4 INTAKE MODELING

C.4.1

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 [128]. 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.

Assumptions

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 IMBA 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 [129]. Table C-8 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-8. Plutonium mixture usage.

Period Plutonium grade

Percentage 239Pu of 

plutonium alpha emitters

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%

10/01/83–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 as it has the 

most favorable to claimant dose conversion factors (ICRP 1994) [130]. 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 1995). Because of the isotopic compositions of the source terms, 

the 234U dose conversion factor yields doses that are favorable to claimants.

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C.4.2

The IMBA computer program was 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.

Bioassay Fitting

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.

C.4.3

For each considered radionuclide or group of radionuclides, the bioassay results were entered into 

IMBA 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 1995) 

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).

Radionuclides and Material Types

The type S 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 people 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

Both type M and S materials were present at Hanford and were common enough to apply to general 

workers, therefore, both types were modeled. This section provides intakes for 239+240Pu of both 

types. Dose reconstructions must also include intakes of other plutonium isotopes and 241Am as 

shown in Table 5-1.

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-9 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.

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Table C-9. Type M 239+240Pu intake periods and rates.

Start Stop

Plutonium intake rate (dpm/d)

50th-percentile 84th-percentile GSD

01/01/1946 12/31/1948 57.7 149.2 2.59

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.

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-10 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-10. Type S 239+240Pu intake period and rates.

Start Stop

Plutonium intake rate (dpm/d)

50th-percentile 84th-percentile GSD

01/01/1946 12/31/1948 1,579 4,094 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.4.3.2 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-17 shows the fit to the 50th-percentile values from all intakes. The figure depicts the 

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-11 summarizes the intake periods and corresponding intake rates 

for the 50th- and 84th-percentile values.

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Table C-11. Type F uranium intake periods and rates.

Start Stop

Uranium intake rate (μg/d)

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-18 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-12 summarizes the 

intake periods and corresponding intake rates for the 50th- and 84th-percentile values.

Table C-12. Type M uranium intake periods and rates.

Start Stop

Uranium intake rate (μg/d)

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-19 to C-24 show the 

individual fits for the 50th-percentile values. The 84th–percentile values were fit similarly. Table C-13 

summarizes the intake rates for the 50th- and 84th-percentile values. Figure C-25 shows the 

predicted excretion rates from all 50th-percentile value type S intakes.

Table C-13. Type S uranium intake periods and rates.

Start Stop

Uranium intake rate (μg/d)

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.4.3.2.4 Strontium-90

Only type F solubility was analyzed for 90Sr. Figure C-26 shows the fit to the 50th-percentile values 

from all intakes. Table C-14 summarizes the intake rates for the 50th- and 84th-percentile values.

Table C-14. Type F 90Sr intake periods and rates.

Start Stop

Sr-90 intake rate (dpm/d)

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.4.3.3 Promethium-147

C.4.3.3.1 Type M

Each chronic intake period for type M 147Pm was fit independently. Figures C-27 and C-28 show the 

individual fits for the 50th-percentile values. The 84th-percentile values were fit similarly. Table C-15 

summarizes the intake rates for the 50th- and 84th-percentile values. Figure C-29 shows the 

predicted excretion rates from all 50th-percentile value type M intakes.

Table C-15. Type M 147Pm intake periods and rates.

Start Stop

147Pm intake rate (dpm/d)

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.4.3.3.2 Type S

Each chronic intake period for type S 147Pm was fit independently. Figures C-30 and C-31 show the 

individual fits for the 50th-percentile values. The 84th–percentile values were fit similarly. Table C-16 

summarizes the intake rates for the 50th- and 84th-percentile values. Figure C-32 shows the 

predicted excretion rates from all 50th-percentile value type S intakes.

Table C-16. Type S 147Pm intake periods and rates.

Start Stop

Pm-147 intake rate (dpm/d)

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.4.3.4 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.

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C.4.3.4.1 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-33 to C-38 show the individual fits for the 50th-percentile 

values. The 84th-percentile values were fit similarly. Table C-17 summarizes the intake rates for the 

50th- and 84th-percentile values. Figure C-39 shows the predicted excretion rates from all 

50th-percentile value type S intakes.

Table C-17. Type S zinc-65 inhalation intake periods and rates. 

Start Stop

Zinc-65 intake rate (pCi/d)

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 – 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.4.3.4.2 Ingestion Intakes

The ingestion intakes of 65Zn were modeled together. Figure C-40 shows the fit to the 50th-percentile 

values from all intakes. The 84th–percentile values were fit similarly. Table C-18 summarizes the 

intake rates for the 50th- and 84th-percentile values. The once-through-cooled reactors were being 

shutdown 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-18. Zinc-65 ingestion intake periods and rates. 

Start Stop

Zn-65 intake rate (pCi/d)

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.4.3.5 Sodium-24

C.4.3.5.1 Inhalation Intakes

Absorption type F was used for 24Na inhalation intakes. Figure C-41 shows the fit to the 

50th-percentile values from all intakes. The 84th–percentile values were fit similarly. Table C-19 

summarizes the intake rates for the 50th- and 84th-percentile values.

C.4.3.5.2 Ingestion Intakes

The ingestion intakes of 24Na were modeled together. Figure C-42 shows the fit to the 50th-percentile 

values from all intakes. These depict the expected excretion rates from an individual who was 

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Table C-19. Type F 24Na inhalation intake periods and rates. 

Start Stop

Na-24 intake rate (pCi/d)

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

exposed for all the periods at the 50th- and 84th-percentile intake rates. Table C-20 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-20. Sodium-24 ingestion intake periods and rates. 

Start Stop

Na-24 intake rate (pCi/d)

50th-percentile 84th-percentile GSD

01/01/1960 09/30/1963 205.6 624.7 3.04

10/01/1963 12/31/1963 1274 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.4.3.6 Cesium-137

Only absorption type F was analyzed for 137Cs. Figure C-43 shows the fit to the 50th-percentile 

values from all intakes. Table C-21 summarizes the intake rates for the 50th- and 84th-percentile 

values.

Table C-21. Type F 137Cs intake periods and rates. 

Start Stop

Cs-137 intake rate (pCi/d)

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-22 to C-31 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 coworker data. Therefore, a GSD of at least 3 was assigned for each of the intake 

periods for the 50th percentile intakes (ORAUT 2007b). The GSDs for different intake periods have 

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also been adjusted in a favorable to claimant manner for consistency between intake periods for 

calculational efficiency. The 95th percentile intakes should be assigned as a constant rather than 

lognormal distribution.

0

1000000

2000000

3000000

4000000

5000000

6000000

1940 1945 1950 1955 1960 1965 1970 1975

Year

Tons U times burnup

Production

Figure C-6. Production rates in tons of uranium times fuel burnup, 1944 to 1972 [131]. 

The bioassay results that were used to establish coworker 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 

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 coworker 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. 

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C.5.1

Dose reconstructors should assume a 12% 239Pu (fuel-grade) mixture for coworker intakes per Table 

C-22. Tables C-22 and C-23 list the 239Pu intakes and associated GSDs for each period for type M 

and S 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 

5.1.2, 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 (72 FR 55214). Therefore, dose 

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

Plutonium

Table C-22. Type M plutonium-239 intake periods and rates. 

Start Stop

Plutonium intake rate (dpm/d)

50th-percentile GSD 95th-percentile

09/01/1946 12/31/1948 57.7 3.00 352

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

1/1/1989 Present 0.1336 8.47 4.49

a. Assumed to be 25 times the 1946 to 1948 rate.

Table C-23. Type S plutonium-239 intake period and rates. 

Start Stop

Plutonium intake rate (dpm/d)

50th-percentile GSD 95th-percentile

09/01/1946 12/31/1948 1579 3.00 9622

01/01/1949 12/31/1952 495.8 3.00 3021

01/01/1953 12/31/1981 12.21 3.18 81.9

01/01/1982 12/31/1988 2.654 7.71 76.4

1/1/1989 Present 2.654 7.71 76.4

a. Assumed to be 25 times the 1946 to 1948 rate.

C.5.2

Tables C-24 to C-26 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. 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/PNWD-3738, Technical Basis Document: Site 

Profiles for Atomic Weapons Employers That Worked Uranium and Thorium Metals (Battelle 2006).

Uranium

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-24 to C-26 from 1952 to the present. The absorption types for the uranium impurities are 

provided in ORAUT-OTIB-0060, Internal Dose Reconstruction (ORAUT 2007b).

C.5.3

Table C-27 lists the 90Sr intakes and associated GSDs for each period. 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 

Strontium-90

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Table C-24. Uranium Type F intake periods and rates. 

Start Stop

Uranium Type F intake (pCi/d)

50th-percentile 

intake GSD

95th-percentile 

intake

12/01/1943 12/31/1948

Not applicable 

(N/A) N/A

N/A

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

1/1/1989 Present 0.446 3.86 4.11

Table C-25. Uranium Type M intake periods and rates. 

Start Stop

Uranium Type M intake (pCi/d)

50th-percentile 

Intake GSD

95th-percentile 

intake

12/01/1943 12/31/1948

See (Battelle 

2006) 5

N/A

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

01/01/1984 12/31/1988 1.75 3.67 14.8

1/1/1989 Present 1.75 3.67 14.8

Table C-26. Uranium Type S intake periods and rates. 

Start Stop

Uranium Type S intake (pCi/d)

50th-percentile 

Intake GSD

95th-percentile 

Intake

12/01/1943 12/31/1948

See (Battelle 

2006) 5 N/A

01/01/1949 12/31/1951 923 3.45 7078

01/01/1952 12/31/1952 1,190 3.45 9126

01/01/1953 12/31/1956 942 3.45 7224

01/01/1957 12/31/1961 1,227 3.45

9409

01/01/1962 12/31/1974 605 3.00 3686

01/01/1975 12/31/1983 76.1 4.82 1012

01/01/1984 12/31/1988 42.6 3.66 361

1/1/1989 Present 42.6 3.66 361

as a contamination source had occurred over 20 years of operation. Air concentration limits were 

essentially unchanged from 1947 to the present (Parker 1947; Patterson 1949). The tolerance air 

concentration in 1945 was 10 times the 1947 value. Therefore, it is reasonable to apply the 90Sr 

intake rates from the 1965 to 1967 period to the 1947 to 1964 period and to apply 10 times those 

rates to the 1944 to 1946 period.

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Table C-27. Type F 90Sr intake periods and rates. 

Start Stop

Sr-90 intake rate (dpm/d)

50th-percentile 

Intake GSD

95th-percentile 

Intake

12/01/1944 12/31/1946 904.7a 3.00 5512

01/01/1947 12/31/1964 90.47b 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

1/1/1989 Present 3.698 3.00 22.5

a. Intake rate assumed to be 10 times the rate for 1965 to 1967.

b. Intake rate assumed to be the same as the rate for 1965 to 1967.

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

Table C-28 lists the 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 2010c). 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 [132]. 

Promethium-147

Table C-28. Promethium-147 intake periods and rates. 

Start Stop

Pm-147 intake (dpm/d)

Type M Type S

50th-percentile 

intake GSD

95th-percentile 

intake

50th-percentile 

intake GSD

95th-percentile 

intake

10/01/1961 12/31/1965a 4,976 3.14 32,684 131,500 3.13 859,204

01/01/1966 12/31/1969 4,976 3.14 32,684 131,500 3.13 859,204

01/01/1970 12/31/1979 1,720 3.00 10,481 34,810 3.00 212,114

a. Intakes set equal to the 1966-69 intakes.

C.5.5

Table C-29 lists the zinc intakes and associated GSDs for each period. 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 1946 period. 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 1947 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 favorable to claimant of the inhalation 

or ingestion intake 

Zinc-65

but not both in the same year. If ingestion is used and the exposure period 

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continues to 1973 or later, then 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.

Table C-29. Zinc-65 intake periods and rates. 

Start Stop

Zn-65 intake (pCi/d)

Inhalation Ingestion

50thpercentile 

intake GSD

95thpercentile 

intake

50thpercentile 

intake GSD

95thpercentile 

intake

12/01/1944 12/31/1946a 445.6 4.07 4484 262 4.01 2573

01/01/1947 12/31/1959b 44.56 4.07 448 26.2 4.01 257

01/01/1960 12/31/1963 44.56 4.07 448 26.2 4.01 257

01/01/1964 12/31/1967 53.68 4.07 540 27.64 4.01 271

01/01/1968 12/31/1970 35.29 3.52 280 16.69 3.44 127

01/01/1971 12/31/1972 29.88 3.52 237 13.06 3.05 82

01/01/1973 12/31/1977 12.58 3.52 99.7 NAc

01/01/1978 12/31/1984 7.411 3.00 45.2 NA

10/1/1976d – 2,555e 5.58 43,212 NA

a. Intake rates assumed to be 10 times the rates for 1960 to 1963.

b. Intake rates assumed to be the same as the rates for 1960 to 1963.

c. NA = not applicable.

d. This is the date of the acute intake.

e. This intake has units of pCi for an acute intake rather than pCi/d.

C.5.6

Table C-30 lists the 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. 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 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 favorable to claimant of the inhalation or ingestion intakes 

Sodium-24

but not 

both in the same year

Table C-30. Sodium-24 intake periods and rates. 

. If ingestion is used and the exposure period continues to 1972 or later, then 

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.

Start Stop

Na-24 intake (pCi/d)

Inhalation Type F Ingestion

50thpercentile 

intake GSD

95thpercentile 

intake

50thpercentile 

intake GSD

95thpercentile 

intake

12/1/1944 12/31/1946a 3,103 3.04 19325 2,056 3.04 12804

01/01/1947 12/31/1959b 310.3 3.04 1932 205.6 3.04 1280

01/01/1960 12/31/1962 310.3 3.04 1932 205.6 3.04 1280

01/01/1963 12/31/1963 1,921 3.61 15872 1,274 3.61 10526

01/01/1964 12/31/1965 1,038 3.52 8228 687.6 3.61 5681

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Start Stop

Na-24 intake (pCi/d)

Inhalation Type F Ingestion

50thpercentile 

intake GSD

95thpercentile 

intake

50thpercentile 

intake GSD

95thpercentile 

intake

01/01/1966 12/31/1971 452.7 3.11 2927 311.9 3.61 2577

01/01/1972 12/31/1973 452.7 3.11 2927 NAc NA

01/01/1974 12/31/1984 250.8 3.11 1622 NA NA

a. Intake rates assumed to be 10 times the rates for 1960 to 1962.

b. Intake rates assumed to be the same as the rates for 1960 to 1962.

c. NA = not applicable.

C.5.7

Table C-31 lists the 137Cs intakes and associated GSDs for each period. Similar to 90Sr, the intake 

rates before 1960 were assumed based on smaller production and similar air concentration limits, with 

the exception of the period from 1944 to 1946.

Cesium-137

Table C-31. Type F cesium-137 intake periods and rates. 

Start Stop

Cs-137 intake rate (pCi/d)

50th-percentile 

intake GSD

95th-percentile 

intake

12/1/1944 12/31/1946 1,186a 3.00 7227

01/01/1947 12/31/1959 118.6b 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 1503

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

1/1/1989 Present 10.13 3.00 61.7

a. Intake rate assumed to be 10 times the rate for 1960.

b. Intake rate assumed to be the same as the rate for 1960.

C.5.8

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 

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 2007a) and for specific buildings in Section 5.6.

Additional Radionuclides

Assignment to unmonitored workers of radionuclides that were not analyzed in this study should be in 

accordance with guidance in Section 5.6.

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STATISTICAL SUMMARIES AND PLOTS

In the following tables, the GM is the 50th-percentile value of the fitted line and GM*GSD is the 

84th-percentile value.

Table C-32. Statistical summary of 239Pu 24-hour urinary excretion rates (dpm/d), 1946 to 1988.

a

Effective 

bioassay 

date

GM 

(50th)

GM*GSD

(84th)

Effective 

bioassay 

date

GM 

(50th)

GM*GSD

(84th)

Effective 

bioassay 

date

GM 

(50th)

GM*GSD

(84th)

07/01/1946 0.25402 0.66092 02/15/1964 0.01826 0.06779 08/15/1976 0.00801 0.02006

07/01/1947 0.20476 0.50991 05/15/1964 0.02133 0.08054 11/15/1976 0.00898 0.02764

04/01/1948 0.21047 0.53721 08/15/1964 0.02435 0.08191 02/15/1977 0.00871 0.02336

10/1/1948 0.22645 0.60437 11/15/1964 0.02067 0.07232 05/15/1977 0.00875 0.02371

07/01/1949 0.10752 0.28933 02/15/1965 0.01409 0.04706 08/15/1977 0.00835 0.02161

07/01/1950 0.10249 0.25412 05/15/1965 0.01380 0.04542 11/15/1977 0.00812 0.02061

07/01/1951 0.10358 0.26263 08/15/1965 0.01248 0.03685 02/15/1978 0.00878 0.02468

07/01/1952 0.05707 0.14968 11/15/1965 0.01556 0.04811 05/15/1978 0.00906 0.02664

04/01/1953 0.01647 0.04436 02/15/1966 0.01197 0.03675 08/15/1978 0.00836 0.02192

10/1/1953 0.01724 0.04873 05/15/1966 0.01152 0.03613 11/15/1978 0.00825 0.02180

02/15/1954 0.01704 0.04637 08/15/1966 0.01354 0.04801 02/15/1979 0.00871 0.02373

05/15/1954 0.01746 0.04932 11/15/1966 0.01200 0.04220 05/15/1979 0.00849 0.02278

08/15/1954 0.01765 0.05080 02/15/1967 0.01330 0.05292 08/15/1979 0.00851 0.02273

11/15/1954 0.01697 0.04609 05/15/1967 0.02204 0.11370 11/15/1979 0.00818 0.02095

02/15/1955 0.01671 0.04478 08/15/1967 0.01175 0.04248 02/15/1980 0.00821 0.02140

05/15/1955 0.01695 0.04624 11/15/1967 0.01338 0.05001 05/15/1980 0.00865 0.02388

08/15/1955 0.01746 0.05021 02/15/1968 0.01216 0.04585 08/15/1980 0.00870 0.02426

11/15/1955 0.01681 0.04538 05/15/1968 0.01078 0.03891 11/15/1980 0.00900 0.02488

02/15/1956 0.01703 0.04723 08/15/1968 0.01397 0.06615 02/15/1981 0.00799 0.02036

05/15/1956 0.01716 0.04803 11/15/1968 0.01192 0.04489 05/15/1981 0.00827 0.02204

08/15/1956 0.01724 0.04828 02/15/1969 0.01321 0.04680 8/5/1981 0.00851 0.02341

11/15/1956 0.01756 0.04956 05/15/1969 0.00952 0.02753 11/5/1981 0.00109 0.00644

02/15/1957 0.01659 0.04369 08/15/1969 0.01101 0.03788 02/15/1982 0.00069 0.00973

05/15/1957 0.01699 0.04599 11/15/1969 0.01095 0.03868 05/15/1982 0.00053 0.00572

08/15/1957 0.01663 0.04360 02/15/1970 0.00974 0.03019 08/15/1982 0.00083 0.00647

11/15/1957 0.01712 0.04671 05/15/1970 0.00888 0.02567 11/15/1982 0.00149 0.01120

02/15/1958 0.01576 0.06661 08/15/1970 0.00980 0.03067 02/15/1983 0.00113 0.00661

05/15/1958 0.01089 0.03745 11/15/1970 0.01307 0.05338 05/15/1983 0.00087 0.00497

08/15/1958 0.01001 0.03106 02/15/1971 0.00933 0.02895 08/15/1983 0.00094 0.00738

11/15/1958 0.01240 0.05037 05/15/1971 0.00993 0.03206 11/15/1983 0.00083 0.00589

02/15/1959 0.01265 0.05288 08/15/1971 0.01126 0.03955 02/15/1984 0.00132 0.00760

05/15/1959 0.01163 0.04018 11/15/1971 0.01089 0.03976 05/15/1984 0.00062 0.01534

08/15/1959 0.01058 0.03276 02/15/1972 0.01179 0.04290 08/15/1984 0.00090 0.00810

11/15/1959 0.01098 0.03388 05/15/1972 0.01013 0.03254 11/15/1984 0.00099 0.00860

02/15/1960 0.01237 0.04317 08/15/1972 0.00973 0.03035 02/15/1985 0.00116 0.01028

05/15/1960 0.01002 0.02983 11/15/1972 0.01073 0.03642 05/15/1985 0.00056 0.00531

08/15/1960 0.00980 0.02880 02/15/1973 0.01107 0.03977 08/15/1985 0.00076 0.00541

11/15/1960 0.00980 0.02931 05/15/1973 0.00948 0.02877 11/15/1985 0.00073 0.00389

02/15/1961 0.00977 0.02866 08/15/1973 0.00945 0.02812 02/15/1986 0.00065 0.00635

05/15/1961 0.00984 0.02971 11/15/1973 0.00875 0.02403 05/15/1986 0.00069 0.01717

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Effective 

bioassay 

date

GM 

(50th)

GM*GSD

(84th)

Effective 

bioassay 

date

GM 

(50th)

GM*GSD

(84th)

Effective 

bioassay 

date

GM 

(50th)

GM*GSD

(84th)

08/15/1961 0.01022 0.03099 02/15/1974 0.00885 0.02454 08/15/1986 0.00099 0.00725

11/15/1961 0.00990 0.02813 05/15/1974 0.00900 0.02532 11/15/1986 0.00050 0.00463

02/15/1962 0.00948 0.02621 08/15/1974 0.00911 0.02602 02/15/1987 0.00069 0.00551

05/15/1962 0.01091 0.03307 11/15/1974 0.01119 0.03852 05/15/1987 0.00065 0.00471

08/15/1962 0.01235 0.04221 02/15/1975 0.00910 0.02516 08/15/1987 0.00086 0.00488

11/15/1962 0.01061 0.03407 05/15/1975 0.00953 0.02738 11/15/1987 0.00048 0.00411

02/15/1963 0.01100 0.03651 08/15/1975 0.00802 0.02030 02/15/1988 0.00080 0.00395

05/15/1963 0.01276 0.04553 11/15/1975 0.00838 0.02233 05/15/1988 0.00052 0.00312

08/15/1963 0.01166 0.03917 02/15/1976 0.00857 0.02433 08/15/1988 0.00064 0.00339

11/15/1963 0.01479 0.05371 05/15/1976 0.00835 0.02245 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. 

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Figure C-8. 50th-percentile plutonium urinalysis data for intakes of Type M 

material, 1949 to 1952.

Figure C-9. 50th-percentile plutonium urinalysis data for intakes of Type M

material, 1953 to 1981.

Figure C-10. 50th-percentile plutonium urinalysis data for intakes of Type M 

material, 1982 to 1988.

● bioassay value

● not used in fit

___ fit

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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.

Figure C-12. 50th-percentile plutonium urinalysis data for intakes of Type S 

material, 1946 to 1948.

● bioassay value

● not used in fit

___ fit

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Figure C-13. 50th-percentile plutonium urinalysis data for intakes of Type S 

material, 1949 to 1952.

Figure C-14. 50th-percentile plutonium urinalysis data for intakes of Type S 

material, 1953 to 1981.

Figure C-15. 50th-percentile plutonium urinalysis data for intakes of Type S 

material, 1982 to 1988.

● bioassay value

● not used in fit

___ fit

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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.

● bioassay value

● not used in fit

___ fit

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Table C-33. Statistical summary of uranium 24-hour urinary excretion rates (μg/d), 1948 to 1988.

a

Effective 

bioassay 

date

GM 

(50th)

GM*GSD

(84th)

Effective 

bioassay 

date

GM 

(50th)

GM*GSD

(84th)

Effective 

bioassay 

date

GM 

(50th)

GM*GSD

(84th)

04/01/1948 3.88 12.48 08/15/1958 4.34 14.70 08/15/1968 2.96 8.82

10/1/1948 5.50 21.41 11/15/1958 3.96 12.54 11/15/1968 3.21 9.10

02/15/1949 4.20 15.22 02/15/1959 3.94 12.97 02/15/1969 2.38 5.98

05/15/1949 3.32 10.87 05/15/1959 4.59 15.29 05/15/1969 2.73 7.06

08/15/1949 7.02 20.71 08/15/1959 5.54 19.18 08/15/1969 2.48 6.26

11/15/1949 4.65 15.47 11/15/1959 4.48 16.23 11/15/1969 2.27 5.67

02/15/1950 6.97 23.35 02/15/1960 4.20 14.36 07/01/1970 2.24 5.65

05/15/1950 4.57 15.20 05/15/1960 5.82 20.97 07/01/1971 2.61 6.88

08/15/1950 4.71 13.26 08/15/1960 4.47 14.79 07/01/1972 3.10 8.19

11/15/1950 3.62 11.81 11/15/1960 4.34 15.45 07/01/1973 2.27 5.34

02/15/1951 3.97 13.61 02/15/1961 4.76 18.96 07/01/1974 2.86 8.96

05/15/1951 4.15 14.93 05/15/1961 4.85 17.54 07/01/1975 0.97 5.99

08/15/1951 4.51 16.53 08/15/1961 3.76 11.26 07/01/1976 0.79 4.84

11/15/1951 4.15 13.99 11/15/1961 3.86 11.32 07/01/1977 0.39 1.42

02/15/1952 5.71 20.24 02/15/1962 3.68 11.21 07/01/1978 0.25 0.76

05/15/1952 4.84 16.17 05/15/1962 3.05 8.70 07/01/1979 0.29 0.96

08/15/1952 4.48 14.71 08/15/1962 2.78 7.73 07/01/1980 0.25 0.74

11/15/1952 5.15 16.08 11/15/1962 2.74 7.69 07/01/1981 0.41 2.02

02/15/1953 4.89 15.49 02/15/1963 3.77 10.83 07/01/1982 0.36 2.11

05/15/1953 4.07 12.44 05/15/1963 3.19 8.55 07/01/1983 0.13 0.64

08/15/1953 3.39 10.92 08/15/1963 2.73 7.43 02/15/1984 0.14 0.82

11/15/1953 3.62 12.79 11/15/1963 2.76 7.37 05/15/1984 0.20 0.78

02/15/1954 3.02 9.77 02/15/1964 2.39 6.25 08/15/1984 0.25 1.15

05/15/1954 3.02 9.26 05/15/1964 2.54 6.79 11/15/1984 0.32 1.34

08/15/1954 3.16 10.11 08/15/1964 2.89 9.30 02/15/1985 0.31 1.25

11/15/1954 3.78 12.74 11/15/1964 2.66 7.54 05/15/1985 0.28 0.89

2/14/1955 3.17 9.87 02/15/1965 3.56 11.55 08/15/1985 0.23 0.71

05/15/1955 3.44 11.06 05/15/1965 3.82 13.26 11/15/1985 0.15 0.45

08/15/1955 2.77 8.14 08/15/1965 4.39 16.82 02/15/1986 0.13 0.38

11/15/1955 3.21 10.35 11/15/1965 3.72 11.94 05/15/1986 0.27 1.36

2/14/1956 2.61 7.45 02/15/1966 3.09 10.22 08/15/1986 0.21 0.78

05/15/1956 3.55 11.48 05/15/1966 3.02 9.28 11/15/1986 0.18 0.62

08/15/1956 4.20 14.94 08/15/1966 3.56 12.10 02/15/1987 0.12 0.43

11/15/1956 2.88 8.92 11/15/1966 3.13 9.35 05/15/1987 0.15 0.50

02/15/1957 4.28 15.51 02/15/1967 2.85 8.54 08/15/1987 0.14 0.42

05/15/1957 5.70 21.26 05/15/1967 2.97 8.84 11/15/1987 0.15 0.43

08/15/1957 6.23 23.63 08/15/1967 2.71 7.71 02/15/1988 0.14 0.41

11/51/1957 6.88 24.31 11/15/1967 2.77 8.13 05/15/1988 0.09 0.23

02/15/1958 6.65 22.18 02/15/1968 3.06 9.14 08/15/1988 0.09 0.29

05/15/1958 5.34 18.07 05/15/1968 3.73 11.62 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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Figure C-17. 50th-percentile uranium urinalysis data for intakes of Type F 

material, 1948 to 1988.

Figure C-18. 50th-percentile uranium urinalysis data for intakes of Type M 

material, 1948 to 1988.

Figure C-19. 50th-percentile uranium urinalysis data for intakes of Type S 

material, 1948 to 1952.

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

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Figure C-20. 50th-percentile uranium urinalysis data for intakes of Type S 

material, 1953 to 1956.

Figure C-21. 50th-percentile uranium urinalysis data for intakes of Type S 

material, 1957 to 1961.

Figure C-22. 50th-percentile uranium urinalysis data for intakes of Type S 

material, 1962 to 1974.

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

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Figure C-23. 50th-percentile uranium urinalysis data for intakes of Type S 

material, 1975 to 1983. 

Figure C-24. 50th-percentile uranium urinalysis data for intakes of Type S 

material, 1984 to 1988. 

Figure C-25. Predicted 50th-percentile urinary excretion of Type S uranium from 

1946 to 1988 based on six independent intakes, compared to bioassay data. 

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

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Table C-34. Statistical summary of 90Sr 24-hour urinary excretion rates (dpm/d), 1966 to 1988.

Effective 

bioassay date GM (50th) GM*GSD (84th)

Effective 

bioassay date GM (50th) GM*GSD (84th)

07/01/1966 20.74 56.46 11/15/1983 0.410 2.73

07/01/1969 14.53 55.88 02/15/1984 0.807 1.76

07/01/1971 1.76 4.56 05/15/1984 1.04 2.70

07/01/1972 2.33 12.27 08/15/1984 1.00 1.66

07/01/1973 1.64 4.51 11/15/1984 1.06 1.87

07/01/1974 1.38 3.82 02/15/1985 1.30 2.88

07/01/1975 0.923 2.57 05/15/1985 1.18 2.38

07/01/1976 1.20 3.40 08/15/1985 1.06 1.69

07/01/1977 0.835 2.12 11/15/1985 1.11 2.20

07/01/1978 1.009 3.82 02/15/1986 1.12 2.26

02/15/1979 0.882 2.55 05/15/1986 1.05 1.88

08/15/1979 2.00 5.58 08/15/1986 1.14 2.05

07/01/1980 1.92 4.92 11/15/1986 1.27 2.29

07/01/1981 2.33 7.60 02/15/1987 1.28 2.02

02/15/1982 0.289 1.43 05/15/1987 1.43 2.28

05/15/1982 0.322 1.58 08/15/1987 1.43 2.18

08/15/1982 0.350 1.72 11/15/1987 1.47 2.21

11/15/1982 0.802 7.91 02/15/1988 1.79 2.79

02/15/1983 0.665 2.60 05/15/1988 1.53 2.57

05/15/1983 0.387 3.13 08/15/1988 1.32 2.02

08/15/1983 0.743 3.78 11/15/1988 0.327 0.993

Figure C-26. 50th-percentile strontium urinalysis data for intakes of Type F 

material, 1965 to 1988.

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● bioassay value

● not used in fit

___ fit
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Table C-35. 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

2/12/1969 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-27. 50th-percentile promethium urinalysis data for intakes of Type M 

material, 1966 to 1969.

Figure C-28. 50th-percentile promethium urinalysis data for intakes of Type M 

material, 1970 to 1979.

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

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Figure C-29. Predicted 50th-percentile urinary excretion of Type M 

promethium from 1966 to 1979 based on two independent intakes, compared 

to bioassay data.

Figure C-30. 50th-percentile promethium urinalysis data for intakes of Type S 

material, 1966 to 1969.

Figure C-31 50th-percentile promethium urinalysis data for intakes of Type S 

material, 1970 to 1979.

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

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Figure C-32. Predicted 50th-percentile urinary excretion of Type S 

promethium from 1966 to 1979 based on two independent intakes, compared 

to bioassay data.

● bioassay value

● not used in fit

___ fit

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Table C-36. Statistical summary of 65Zn measured in whole-body counts (nCi), 1960 to 1984. 

Effective 

bioassay 

date

GM 

(50th)

GM*GSD 

(84th)

Effective 

bioassay 

date

GM 

(50th)

GM*GSD 

(84th)

Effective 

bioassay 

date

GM 

(50th)

GM*GSD 

(84th)

07/01/1960 2.45 7.80 05/15/1969 1.28 4.63 05/15/1977 0.788 2.83

04/01/1961 2.16 9.58 08/15/1969 1.05 3.66 08/15/1977 0.688 2.67

10/1/1961 2.13 7.78 11/15/1969 1.45 5.37 11/15/1977 0.388 1.27

04/01/1962 2.04 8.73 02/15/1970 1.45 5.56 02/15/1978 0.371 1.16

10/1/1962 2.15 12.08 05/15/1970 1.27 4.31 05/15/1978 0.369 1.15

02/15/1963 2.11 8.44 08/15/1970 1.56 5.11 08/15/1978 0.373 1.19

05/15/1963 1.60 6.25 11/15/1970 2.10 5.87 11/15/1978 0.396 1.28

08/15/1963 1.34 4.70 02/15/1971 1.52 5.06 02/15/1979 0.401 1.28

11/15/1963 3.40 12.52 05/15/1971 1.55 4.69 05/15/1979 0.415 1.37

02/15/1964 2.40 9.02 08/15/1971 1.29 3.81 08/15/1979 0.348 1.02

05/15/1964 2.12 7.22 11/15/1971 1.24 3.55 11/15/1979 0.406 1.31

08/15/1964 1.92 6.79 02/15/1972 1.18 3.92 02/15/1980 0.384 1.19

11/15/1964 4.20 16.06 05/15/1972 0.963 3.01 05/15/1980 0.403 1.32

02/15/1965 2.93 12.17 08/15/1972 1.17 3.74 08/15/1980 0.372 1.15

05/15/1965 3.11 10.80 11/15/1972 1.18 3.98 11/15/1980 0.348 1.01

08/15/1965 2.348 9.40 02/15/1973 1.18 4.21 02/15/1981 0.372 1.09

11/15/1965 2.56 10.10 05/15/1973 1.04 3.44 05/15/1981 0.354 1.04

02/15/1966 2.82 9.32 08/15/1973 0.612 1.99 08/15/1981 0.306 0.817

05/15/1966 2.87 10.10 11/15/1973 0.678 2.27 11/15/1981 0.316 0.854

08/15/1966 1.84 6.32 02/15/1974 0.497 1.61 02/15/1982 0.340 0.947

11/15/1966 1.86 8.03 05/15/1974 0.431 1.33 05/15/1982 0.338 0.944

02/15/1967 1.78 6.06 08/15/1974 0.456 1.41 08/15/1982 0.334 0.934

05/15/1967 3.35 12.02 11/15/1974 0.507 1.64 11/15/1982 0.331 0.909

08/15/1967 2.13 8.26 04/01/1975 0.405 1.23 02/15/1983 0.323 0.884

11/15/1967 1.50 5.91 10/1/1975 0.613 2.00 05/15/1983 0.334 0.922

02/15/1968 1.53 6.61 02/15/1976 0.647 2.19 08/15/1983 0.314 0.843

05/15/1968 1.38 5.45 05/15/1976 0.712 3.25 11/15/1983 0.337 0.937

08/15/1968 1.41 6.18 08/15/1976 0.673 2.46 04/01/1984 0.354 1.03

11/15/1968 1.83 6.78 11/15/1976 1.21 5.39 10/01/1984 0.391 1.09

02/15/1969 2.35 7.89 02/15/1977 0.964 3.74

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Figure C-33. 50th-percentile zinc whole-body counting data for inhalation 

intakes of Type S material, 1960 to 1963.

Figure C-34. 50th-percentile zinc whole-body counting data for inhalation 

intakes of Type S material, 1964 to 1967.

Figure C-35. 50th-percentile zinc whole-body counting data for inhalation 

intakes of Type S material, 1968 to 1970.

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

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Figure C-36. 50th-percentile zinc whole-body counting data for inhalation 

intakes of Type S material, 1971 to 1972.

Figure C-37. 50th-percentile zinc whole-body counting data for inhalation 

intakes of Type S material, 1973 to 1977.

Figure C-38. 50th-percentile zinc whole-body counting data for inhalation 

intakes of Type S material, 1978 to 1984.

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

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Figure C-39. 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. 

Figure C-40. 50th-percentile zinc whole-body counting data for ingestion 

intakes, 1960 to 1984. 

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

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Table C-37. Statistical summary of 24Na measured in whole-body counts (nCi), 1960 to 1984. 

Effective 

bioassay 

date

GM 

(50th)

GM*GSD 

(84th)

Effective 

bioassay 

date

GM 

(50th)

GM*GSD 

(84th)

Effective 

bioassay 

date

GM 

(50th)

GM*GSD 

(84th)

07/01/1960 0.221 0.587 02/15/1969 0.194 0.570 11/15/1976 0.158 0.434

04/01/1961 0.174 0.610 05/15/1969 0.204 0.649 02/15/1977 0.168 0.465

10/1/1961 0.187 0.616 08/15/1969 0.268 0.915 05/15/1977 0.166 0.466

02/15/1962 0.142 0.421 11/15/1969 0.285 0.979 08/15/1977 0.162 0.449

05/15/1962 0.182 0.609 02/15/1970 0.218 0.623 11/15/1977 0.162 0.458

08/15/1962 0.150 0.413 05/15/1970 0.228 0.605 02/15/1978 0.138 0.368

11/15/1962 0.145 0.414 08/15/1970 0.193 0.543 05/15/1978 0.148 0.401

02/15/1963 0.180 0.562 11/15/1970 0.196 0.559 08/15/1978 0.139 0.368

05/15/1963 0.185 0.515 02/15/1971 0.260 0.711 11/15/1978 0.150 0.414

08/15/1963 0.231 0.713 05/15/1971 0.234 0.591 02/15/1979 0.134 0.356

11/15/1963 1.11 4.02 08/15/1971 0.224 0.589 05/15/1979 0.155 0.421

02/15/1964 0.388 1.35 11/15/1971 0.220 0.591 08/15/1979 0.147 0.396

05/15/1964 0.355 1.16 02/15/1972 0.245 0.715 11/15/1979 0.145 0.387

08/15/1964 0.316 0.993 05/15/1972 0.255 0.760 02/15/1980 0.139 0.369

11/15/1964 0.647 2.06 08/15/1972 0.206 0.608 05/15/1980 0.140 0.372

02/15/1965 0.555 2.17 11/15/1972 0.243 0.673 08/15/1980 0.131 0.346

05/15/1965 0.295 0.983 02/15/1973 0.244 0.756 11/15/1980 0.140 0.370

08/15/1965 0.286 0.930 05/15/1973 0.196 0.556 02/15/1981 0.139 0.372

11/15/1965 0.292 0.891 08/15/1973 0.220 0.614 05/15/1981 0.138 0.366

02/15/1966 0.347 1.09 11/15/1973 0.176 0.501 08/15/1981 0.147 0.403

05/15/1966 0.233 0.650 02/15/1974 0.144 0.401 11/15/1981 0.133 0.348

08/15/1966 0.257 0.834 05/15/1974 0.147 0.431 02/15/1982 0.134 0.352

11/15/1966 0.218 0.687 08/15/1974 0.149 0.418 05/15/1982 0.149 0.407

02/15/1967 0.288 0.904 11/15/1974 0.139 0.369 08/15/1982 0.140 0.375

05/15/1967 0.428 1.740 02/15/1975 0.128 0.331 11/15/1982 0.143 0.387

08/15/1967 0.247 0.701 05/15/1975 0.131 0.340 02/15/1983 0.145 0.399

11/15/1967 0.346 0.998 08/15/1975 0.140 0.383 05/15/1983 0.142 0.385

02/15/1968 0.405 1.610 11/15/1975 0.131 0.342 08/15/1983 0.142 0.390

05/15/1968 0.269 0.887 02/15/1976 0.139 0.371 11/15/1983 0.150 0.414

08/15/1968 0.229 0.728 05/15/1976 0.148 0.404 04/01/1984 0.144 0.377

11/15/1968 0.256 0.805 08/15/1976 0.166 0.459 10/01/1984 0.171 0.495

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Figure C-41. 50th-percentile sodium whole-body counting data for inhalation 

intakes of Type F material, 1960 to 1984.

Figure C-42. 50th-percentile sodium whole-body counting data for ingestion 

intakes, 1960 to 1984.

● bioassay value

● not used in fit

___ fit

● bioassay value

● not used in fit

___ fit

ATTACHMENT C

INTERNAL DOSIMETRY COWORKER DATA FOR THE HANFORD SITE

Page 52 of 54
Document No. ORAUT-TKBD-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 162 of 163

Table C-38. Statistical summary of 137Cs measured in whole-body counts (nCi), 1960 to 1988. 

Effective 

bioassay 

date

GM 

(50th)

GM*GSD 

(84th)

Effective 

bioassay 

date

GM 

(50th)

GM*GSD 

(84th)

Effective 

bioassay 

date

GM 

(50th)

GM*GSD 

(84th)

02/15/1960 6.773 9.620 02/15/1969 11.964 18.931 02/15/1979 1.1775 2.9579

05/15/1960 6.689 9.347 05/15/1969 10.544 16.593 05/15/1979 0.8671 2.0518

08/15/1960 5.977 9.466 08/15/1969 8.526 13.837 08/15/1979 0.9219 2.1979

11/15/1960 5.770 8.302 11/15/1969 8.198 13.210 11/15/1979 1.2393 2.6536

02/15/1961 4.728 6.858 02/15/1970 6.417 12.039 02/15/1980 1.1519 2.8199

05/15/1961 4.482 6.764 05/15/1970 5.877 9.856 05/15/1980 0.8175 2.9124

08/15/1961 3.805 6.006 08/15/1970 5.417 9.242 08/15/1980 0.6695 1.8971

11/15/1961 3.491 5.184 11/15/1970 4.770 8.025 11/15/1980 0.7321 1.9398

02/15/1961 3.527 5.563 02/15/1971 4.049 7.580 02/15/1981 0.8278 2.1018

05/15/1961 3.457 5.055 05/15/1971 3.195 6.012 05/15/1981 0.4992 1.1927

08/15/1961 4.378 6.502 08/15/1971 2.574 4.939 08/15/1981 0.3353 0.9534

11/15/1961 5.253 7.678 11/15/1971 2.667 5.467 11/15/1981 0.4740 1.0964

02/15/1963 6.691 10.362 02/15/1972 2.824 5.511 02/15/1982 0.5655 1.3484

05/15/1963 6.773 9.620 05/15/1972 3.022 5.516 05/15/1982 0.4695 1.2468

08/15/1963 6.689 9.347 08/15/1972 2.176 4.728 08/15/1982 0.4364 1.0269

11/15/1963 5.977 9.466 11/15/1972 2.222 4.332 11/15/1982 0.5316 1.2952

02/15/1964 5.770 8.302 02/15/1973 2.0544 4.5035 02/15/1983 0.4257 1.1238

05/15/1964 4.728 6.858 05/15/1973 1.8482 3.4682 05/15/1983 0.3023 0.8154

08/15/1964 4.482 6.764 08/15/1973 1.8296 3.3453 08/15/1983 0.3705 0.9053

11/15/1964 3.805 6.006 11/15/1973 1.6404 3.3004 11/15/1983 0.4354 1.1195

02/15/1965 3.491 5.184 02/15/1974 1.2505 3.0204 02/15/1984 0.2487 0.6537

05/15/1965 3.527 5.563 05/15/1974 0.9186 2.1838 05/15/1984 0.2461 0.6285

08/15/1965 3.457 5.055 08/15/1974 1.0889 2.4621 08/15/1984 0.2430 0.6134

11/15/1965 4.378 6.502 11/15/1974 1.1091 2.7017 11/15/1984 0.2398 0.5927

02/15/1966 5.253 7.678 02/15/1976 1.0488 2.1653 02/15/1985 0.2450 0.6327

05/15/1966 6.691 10.362 05/15/1976 1.1197 2.6952 05/15/1985 1.1416 2.7971

08/15/1966 8.922 13.998 08/15/1976 0.9322 2.1818 08/15/1985 1.1220 2.8167

11/15/1966 11.218 16.663 11/15/1976 1.6176 3.9736 11/15/1985 1.1235 2.8170

02/15/1967 13.470 19.952 02/15/1977 1.3651 2.9877 04/01/1986 1.1169 2.7720

05/15/1967 15.374 23.543 05/15/1977 1.4688 3.4256 10/1/1986 1.1106 2.7455

08/15/1967 17.644 26.857 08/15/1977 1.1219 2.9244 04/01/1987 0.9186 2.1838

11/15/1967 18.129 26.767 11/15/1977 0.8683 2.2466 10/01/1987 1.0889 2.4621

02/15/1968 17.811 28.334 02/15/1978 1.0525 2.6430 04/01/1988 1.1091 2.7017

05/15/1968 15.208 23.287 05/15/1978 0.7876 2.3797 10/01/1988 1.0488 2.1653

08/15/1968 15.288 23.190 08/15/1978 0.7788 2.2337

11/15/1968 13.341 19.881 11/15/1978 1.0824 2.6808

ATTACHMENT C

INTERNAL DOSIMETRY COWORKER DATA FOR THE HANFORD SITE

Page 53 of 54
Document No. ORAUT-TKBD-0006-5 Revision No. 04 Effective Date: 10/20/2010 Page 163 of 163

Figure C-43. 50th-percentile 137Cs whole-body counting data for inhalation 

intakes of Type F material, 1960 to 1988. 

SITE RESEARCH DATABASE REFERENCE NUMBERS

Table C-39. Hanford monthly reports. 

358, 434, 442, 451, 5044, 5047, 5047, 5054, 5055, 5057, 5060, 5072, 5087, 5090, 5098, 5199, 5199, 

33220, 33226, 33226, 33228, 33265, 33266, 33268, 33271, 34248, 34251, 34254, 34255, 34267, 

34270, 34270, 34271, 34272, 34272, 34273, 34273, 34281, 34285, 34285, 34290, 34293, 34293, 

34294, 34302, 34304, 34307, 34308, 34309, 34310, 34341, 34350, 34355, 34604, 35862, 35863, 

35863, 35864, 35864, 35866, 35867, 36737, 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, 37089, 37229, 37234, 37239, 37244, 37246, 37248, 

37256, 37261, 37265, 37270, 37274, 37280, 37280, 37281, 37282, 37297, 37306, 37311, 37323, 

37334, 37341, 37346, 37364, 37374, 37390, 37400, 37405, 37409, 37411, 37418, 37422, 37435, 

37435, 37440, 37440, 37474, 37556, 37559, 37561, 37563, 37564, 37568, 37569, 37571, 37580, 

37580, 37582, 37586, 37587, 37591, 40711, 40711, 40712, 40712, 40726, 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, 

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, 

41388, 51245, 52690, 52693, 53051, 53364, 53366, 53367

Table C-40. 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, 

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

● bioassay value

● not used in fit

___ fit

ATTACHMENT C

INTERNAL DOSIMETRY COWORKER DATA FOR THE HANFORD SITE

Page 54 of 54