ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 89

Document Title:

Hanford Site – Occupational Internal Dose

Document Number: ORAUT-TKBS-0006-5 

Revision: 02

Effective Date: 06/22/2007

Type of Document TBD

Supersedes: Revision 01

Subject Expert(s): Edward D. Scalsky

Site Expert(s): Donald E. Bihl

Approval: Signature on File Approval Date: 06/04/2007

Edward D. Scalsky, Document Owner

Concurrence: Signature on File Concurrence Date: 05/23/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 05/15/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 05/17/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 06/22/2007

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. 02 Effective Date: 06/22/2007 Page 2 of 89

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 Pu-239 contaminant level in RU to 0.8 nCi/gU; added 

information about U-233 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 info. about pure Pu-238 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 Am-241 recovery 

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

Cm-244 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 U-233 production. Sec. 5.2.8: 

added some info. about promethium use. Sec. 5.2.9: removed 

discussion about background levels of Po-210. Sec. 5.2.10: added 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 3 of 89

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

some info. about neptunium urinalyses. Sec. 5.2.11: added info. 

about thorium urinalyses and S-35 urinalyses. Sec. 5.3.1: revised 

Table 5.3.1-1 (5-22) to clarify MDAs, decision levels and reporting 

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

thorium monitoring. Sec. 5.4: replaced the old mixture rules with use 

of Table 5.7-2 (5-25); Sec. 5.6.2: (chronic intakes) deleted entire 

section. Sec. 5.7 (now 5.6): added information about tritium in the 

400 area drinking water; added reference to OTIB-0039 and revised 

Table 5-25 accordingly. Moved old Appendix D into this document as 

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

a total rewrite of the document. Incorporates NIOSH formal review 

comments. Many grammatical and format changes throughout. 

Table 5-12: removed the uranium specific activity data for DU and 

natural U; added statement to use IMBA defaults to footnote a. 

Section 5.3.4: added table showing buildings with thorium exposure; 

added text and table explaining adjustment factors for whole body 

measurements of Th-232 and associated intakes; added thoron 

doses for workers processing thorium in 3722 and 3732 bldgs. 

Section 5.6: clarified possible absorption types for Pu and U in the 

specific-facilities section; added thoron to Table 5-30; revised tritium 

intakes for unmonitored workers due a coworker analysis of tritium 

doses. New Tables 5-27 and 5-28 providing Th-232 intakes based on 

whole body counting at long times after end of chronic intakes. 

Section 5.3.4: revised thoron WLMs, removed reference to thoron 

dose with a reference to TIB-011, clarified use of OTIB-0039 for 

unmonitored thorium intakes. Section 5.6: removed references to 6-

mrem tritium dose and stayed with intake values; clarified how the 

tritium intake values were determined for the 1955-60, 1970-71, and 

1973-83 time periods; corrected a math error resulting in slightly 

increased intakes for 1984, 85, and 86. Includes Attributions and 

Annotations and additional references. Incorporates formal internal 

review comments for Attributions and Annotations section. This 

revision results in an increase in assigned dose and a PER is 

required. Training required: As determined by the Task Manager. 

Initiated by Donald E. Bihl and Edward D. Scalsky.
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 4 of 89

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

7

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

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

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

5.2 In Vitro Minimum Detectable Activities, Analytical Methods, and 

Reporting Protocols................................................................................................................14

5.2.1 Plutonium .................................................................................................................15

5.2.2 Americium ................................................................................................................20

5.2.3 Curium......................................................................................................................22

5.2.4 Tritium ......................................................................................................................22

5.2.5 Uranium....................................................................................................................24

5.2.6 Fission Product Analysis...........................................................................................29

5.2.7 Strontium..................................................................................................................30

5.2.8 Promethium ..............................................................................................................32

5.2.9 Polonium ..................................................................................................................33

5.2.10 Neptunium................................................................................................................34

5.2.11 Other Limited-Exposure Radionuclides.....................................................................34

5.3 In Vivo Minimum Detectable Activities, Analytical Methods, and 

Reporting Protocols................................................................................................................35

5.3.1 Whole-Body Counters...............................................................................................35

5.3.2 Chest Counters.........................................................................................................39

5.3.3 Thyroid Counters ......................................................................................................42

5.3.4 Thorium Exposure and Monitoring............................................................................43

5.3.5 Head Counters and Other Counts.............................................................................47

5.3.6 General Notes About Items in the Database.............................................................47

5.4 Mixtures ................................................................................................................................. 48

5.4.1 Fission and Activation Products................................................................................49

5.5 Interferences, Uncertainties ................................................................................................... 50

5.5.1 Contamination of Samples........................................................................................50

5.5.2 Uncertainties ............................................................................................................50

5.6 Unmonitored Workers ............................................................................................................ 51

5.7 Attributions and Annotations .................................................................................................. 58

References ........................................................................................................................................71

Glossary ............................................................................................................................................80
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 5 of 89

ATTACHMENT A, CODES USED IN BIOASSAY AND TOLERANCE VALUES ................................ 81

A.1 Codes Used in Bioassay and Internal Dose Records .............................................................82

A.2 Tolerance Dose and Tolerance Values at Hanford................................................................. 87

LIST OF TABLES

TABLE TITLE

5-1 Codes and radionuclides associated with bioassay at Hanford ..............................................

PAGE

12

5-2 Routine plutonium urinalysis detection levels ......................................................................... 16

5-3 MDAs for nonroutine Pu excreta analyses ............................................................................. 17

5-4 Activity composition of Hanford reference weapons-grade plutonium 

mixture (6%)...........................................................................................................................18

5-5 Activity composition of Hanford reference fuel-grade plutonium mixture 

(12%).....................................................................................................................................18

5-6 Activity composition of Hanford reference commercial power fuel-grade 

plutonium mixture................................................................................................................... 19

5-7 Routine 241Am urinalysis detection levels ............................................................................... 21

5-8 MDAs for nonroutine 241Am excreta analyses......................................................................... 22

5-9 Routine Cm urinalysis detection levels ................................................................................... 23

5-10 MDAs for nonroutine Cm excreta analyses ............................................................................ 23

5-11 Routine tritium urinalysis detection levels............................................................................... 24

5-12 Radiological characteristics of Hanford uranium mixtures ...................................................... 25

5-13 Impurities in recycled uranium at Hanford .............................................................................. 25

5-14 Inhalation class for Hanford uranium compounds................................................................... 26

5-15 Routine uranium urinalysis detection levels............................................................................ 27

5-16 MDAs for nonroutine uranium excreta analyses ..................................................................... 28

5-17 Routine fission product urinalysis detection levels.................................................................. 30

5-18 Routine 90Sr urinalysis detection levels .................................................................................. 31

5-19 Routine 147Pm urinalysis detection levels ............................................................................... 32

5-20 Routine 210Po urinalysis detection levels ................................................................................ 33

5-21 MDAs for nonroutine 210Po excreta analyses.......................................................................... 33

5-22 Routine whole-body counting detection levels........................................................................ 36

5-23 Mean body burdens of 137Cs from fallout in the United States ................................................ 40

5-24 Routine chest counting detection levels ................................................................................. 41

5-25 Buildings with thorium processing .......................................................................................... 43

5-26 232Th intakes from whole body count measurements.............................................................. 45

5-27 232Th to 228Ac ratios in the whole body after end of intake period, type M ............................... 46

5-28 232Th to 228Ac ratios in the whole body after end of intake period, type S................................ 46

5-29 Statistical parameters of Hanford recorded tritium doses and associated 

intakes ...................................................................................................................................53

5-30 Intakes for workers with external monitoring but essentially no bioassay ............................... 56

A-1 Sample type codes................................................................................................................. 82

A-2 Bioassay reason codes .......................................................................................................... 82

A-3 Excreta sample kit codes ....................................................................................................... 83

A-4 In vivo body location codes .................................................................................................... 84

A-5 Excreta unit codes..................................................................................................................84

A-6 Excreta processing codes ...................................................................................................... 84
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 6 of 89

A-7 Excreta laboratory codes........................................................................................................ 85

A-8 Excreta no-sample codes....................................................................................................... 85

5-9 In vivo invalid result codes ..................................................................................................... 85

A-10 INTERTRAC mode-of-intake codes ....................................................................................... 86

A-11 INTERTRAC evaluation reason codes ................................................................................... 86

A-12 INTERTRAC source-of-intake codes...................................................................................... 86

A-13 INTERTRAC miscellaneous codes......................................................................................... 86

A-14 Special whole-body count resolution codes............................................................................ 87
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 7 of 89

ACRONYMS AND ABBREVIATIONS

BPNNL Battelle - Pacific Northwest National Laboratory

CADW Chronic Annual Dose Workbook

CF commercial fuel

DAC derived air concentration

DIL derived investigation level

DOE U.S. Department of Energy

DOE-RL DOE Richland Operations Office

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ft feet

GeLi lithium-drifted germanium

GI gastrointestinal

GSD geometric standard deviation

HIE Hanford Internal Exposure (database)

HPGe high-purity germanium

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Assessment

KPA kinetic phosphorescence analysis

LANL Los Alamos National Laboratory

LEPD Low Energy Photon Detector (also computer code to indicate use of the LEPD)

MDA minimum detectable activity or, for elemental uranium, minimum detectable amount

MDC minimum detectable concentration 

min minute

MPBB maximum permissible body burden

MPC maximum permissible concentration

NBS National Bureau of Standards

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)

PNL Pacific Northwest Laboratory

PNNL Pacific Northwest National Laboratory

POC probability of causation

PRTR Plutonium Recycle Test Reactor
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 8 of 89

RDA Reliably Detectable Activity

REX Radiological Exposure (database)

RL Richland Operations Office

RU recycled uranium

SRDB Ref ID Site Research Database Reference Identification (number)

TPU total propagated uncertainty

TTA thenoyl trifluoroacetone

U.S.C. United States Code

UST United States Testing Company
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 9 of 89

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

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

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

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

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

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

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

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

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

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

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

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

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

derived:

] guidelines established under 

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

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

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

• Radiation from naturally occurring radon present in conventional structures

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

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 10 of 89

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 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

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 NIOSH IREP (Interactive 

RadioEpidemiological Program) computer code. 

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 was 

operating three reactors, a fuel manufacturing facility 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. The coworker analysis document, Internal Dosimetry Coworker Data for the 

Hanford Site, OTIB-0039 (ORAUT 2007), provides guidance on assigning intakes for the years prior 

to 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/day from 

alpha-emitters, 1 rep/day to the thyroid from beta/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/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 compiled by R. H. Wilson (1987), one of the priority tasks for a special studies 

group 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 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 11 of 89

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 prior to 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,” etc. 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 used to 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. 

Table 5-1 provides a fairly exhaustive list of codes for analyses that dose reconstructors might 

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

used for scheduling bioassay but not for reporting 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 

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

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

X-ray/gamma-ray analysis on an excreta sample using the low-energy photon detector (a thin-window 

germanium detector); however, if anything was detected, the actual radionuclide was reported [2]. 

The code GOK shows on in vivo count hardcopy 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 NaIbased 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/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 pertaining to the nature of the intake, including:

• Reason for an intake assignment,

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

• Nature of intake, and

• Mode of intake.

Tables listing and explaining these codes are provided in Attachment A.
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 12 of 89

Table 5-1. Codes and radionuclides associated with bioassay at Hanford [4]. 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

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

IPUL Low-level isotopic Pu Pu-238 and Pu-239 using a 10,000-minute count
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 13 of 89

Code Description Comment

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/U-natural 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

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 Natural uranium

U 233 Uranium-233 See uranium discussion in text
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 14 of 89

Code Description Comment

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

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 the electronic database(s). However, 

for some of the earliest urinalysis records, cases have been discovered where not all records were 

included in the electronic database. For any case where urinalysis might have been obtained prior to 

1974, the hardcopy file for the case should be thoroughly reviewed for urinalysis results that might be 

missing in the electronic database. The Hanford Internal Exposure (HIE) database was implemented 

in 1974, 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 could be isolated situations where 

some data was never entered in a database or did not get transferred from one database to another

[5]. 

There is another anomaly found in the results from around 1946-1950. 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; however, experience has shown that this convention was also used to indicate a result that was 

a non-detection. 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

[6]. 

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 used most frequently (kit code 1) was a simulated 24-hr sample 

obtained by sampling from evening through morning on two consecutive nights (or equivalence for 

shift workers). Termination samples were often simulated 12-hr samples (kit code 2) [7]. The dose 

reconstructor needs to 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) in 1965. At that time, the U.S. Department of Energy (DOE) Richland Office established a 

contract for in vitro analyses with the United States Testing Company (UST), which built and operated 

a commercial low-level radiochemistry laboratory in north Richland until 1990. The responsibility for 

awarding and overseeing the contract was subsequently transferred to Battelle as operators of the 

Pacific Northwest Laboratory (PNL). 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: United States Testing, International 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 15 of 89

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

The current in vitro laboratory is General Engineering Laboratory [8]. 

Battelle defaulted 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/ 

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, with the last 

results received in March 1992. This interruption in the normal process of bioassay monitoring had 

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

detectable activities 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

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 combined with nuclear track emulsion 

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

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

levels in Table 5-2 fit reasonably with the concept of limit of detection or Minimum Detectable Activity 

(MDA). For example, Wilson (1987) states, 

Plutonium

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.

Prior to 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 [9]. 

The results might have been reported as Pu 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 dpm/sample, but the same results were recorded in units of μCi/sample in the electronic 

database. The units in the electronic 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 (onehalf of the nominal 0.05 dpm MDA). This method of recording was used through 1974. In 1975, the 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 16 of 89

Table 5-2. Routine plutonium urinalysis detection levelsa. 

Period

MDA, 

dpm/sample

Decision level, 

dpm/sample Measured quantity

Prior to June 1949 0.96b 0.66 Total Pu alpha

6/1949 to 11/1952 0.33 Total Pu alpha

12/1952 to 1/27/53 0.18 Total Pu alpha

1/28/53 to 3/26/53 0.15 Total Pu alpha

3/27/53 to 11/06/53 0.05 Total Pu alpha

11/07/53 to 12/04/53 0.07 Total Pu alpha

12/53 to 4/55 0.057 Total Pu alpha

5/55 to 8/55 0.027c Total Pu alpha

9/55 to 9/55 0.04c Total Pu alpha

10/55 to 9/30/83 0.05d 0.025d Total Pu alpha

10/01/83 to 12/31/83 0.035 Not established, but 

0.018 is reasonable

Each Pu-238, Pu239

1/02/84 to 4/88 0.02 Total Pu alpha

5/88 to 5/90 0.02 0.01 Total Pu alpha

6/90 to 11/91 0.03 0.015 Total Pu alpha

11/91 to 4/2000 0.02 0.01 Total Pu alpha

5/2000 to 8/2001 0.02 Xb + 2.05 × TPUe Total Pu alpha

9/2001 to 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. Recommend using 

0.05 dpm as the MDA [11]. 

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 [12].

e. Xb is mean of blanks and TPU is total propagated uncertainty. The Xb values used 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 [13].

units were changed to dpm/sample (unit code 1) and 0.025 was recorded for results for which 

plutonium was not detected [10]. 

In October 1983 several changes were made. The lanthanum fluoride/TTA method was replaced by 

the use of anion exchange columns, alpha spectrometry analysis replaced autoradiography, 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-min count time has been used 

since 1984. A 10,000-min count time was introduced for special situations in 1996 but its use was 

rare [14]. 

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 [15]. 

Fecal samples were usually not analyzed in total (were aliquoted after muffling, dry ashing, and wet 

ashing); hence, more than one analysis result for a given sample was possible and will often be found 

in the database [16]. 

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

general, the laboratory performed slightly better than the contractual MDA, but the true MDA varied 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 17 of 89

slightly over time and the contractual MDA was a reliable estimate [17]. Reporting of errors, which 

was the total propagated uncertainty including uncertainty associated with the determination of 

chemical yield, counting efficiency determination, and systematic errors, began in 1981 [18]. 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 was used initially 

for all results, being one half the nominal MDA. The decision level was allowed to become samplespecific based on the total propagated uncertainty in 2000, and an adjustment was made to the 

formula in 2001 (Lynch et al. 2000 and 2001). 

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

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

(emergency and expedite) have been available for special urine and fecal samples. The contractual 

MDAs for these samples are provided in Table 5-3.

Table 5-3. MDAs for nonroutine Pu excreta analyses [19]. 

Period

Fecal samples, MDA, dpm/sample Urine samples, MDA, dpm/sample

Emergencya Expedite Priority Emergencya Expedite

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

10/1983 to 1/1985c 9 NA 0.2 0.5 NA

1/1985 to 6/1990 9 3 0.2 0.5 0.08

6/1990 to 2/1991d 20 4 NA 2 0.4

2/1991 to present 9 3 0.2 0.5 0.08

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

b. MDAs varied according to sample size over the range shown; the lower value was generally applicable except 

for very large samples (Larson 1982). MDAs for this period apply to total Pu 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 PNNL’s 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; however, routine fecal sampling 

was used for some high-risk plutonium workers, mostly operators at PUREX 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, initiated in response to the study, showed frequent-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 [20], 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 differing 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, 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 18 of 89

the target mixture was the weapons-grade mixture [21]. There is evidence at Los Alamos National 

Laboratory that plutonium produced at Hanford in the 1940s had less 240Pu, perhaps closer to 3% 

(ORAUT 2004); however, documentation of that at Hanford has not been found. Tables 5-4 and 5-5 

list the relative activities of plutonium isotopes and 241Am, which grows in from 241Pu, for 6% 240Pu and 

12% 240Pu mixtures (from Carbaugh 2003a). In these tables, “aging” refers to the time since the 

241Am was separated from the plutonium then starts to build in again from decay of 241Pu.

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

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

Table 5-4. Activity composition of Hanford reference weapons-grade plutonium mixture (6%). 

Mixture designation: Fresh 5-year 10-year 15-year 20-year 25-year 30-year

Years of aginga: 0 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 0 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 NA 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 NA 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.

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

Mixture designation: Fresh 5-year 10-year 15-year 20-year 25-year 30-year

Years of aginga: 0 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 0 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 NA 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 NA 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.
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 19 of 89

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. 

Prior to 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 handled at Hanford were nominally weapons grade and, if the 239+240Pu-to- 238Pu ratio implies weapons grade (e.g., >8), dose reconstructors should use the ratios in Table 5-4. 

However, if no information is available about the nature of the plutonium mixture associated with an 

intake, an assumption of 10-year-old fuel-grade plutonium mixture would be favorable to claimants 

and reasonable, except it is recommended to assume fresh plutonium for the 1944 through 1949 and 

5-year-old plutonium for 1950 through 1954. For intakes since about 1996, 20-year-old fuel-grade 

mixture could be assumed. 

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

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

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

higher degree of “burn-up,” and those at Hanford were characterized by much more 241Pu and 

nominally 26% 240Pu. Table 5-6 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. This material would be rich in 241Am. Plutonium from the West 

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

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

default mixtures mentioned above should be used.

Table 5-6. Activity composition of Hanford reference commercial power fuel-grade plutonium mixture

(Carbaugh 2003a). 

Mixture designation: Fresh 5-year 10-year 15-year 20-year 25-year 30-year

Years of aginga: 0 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 0 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 NA 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 NA 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 mix.

If some of the plutonium bioassay was obtained prior to October 1983 and some after, the two data 

sets are not compatible; the prior-to-October-1983 results 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 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 20 of 89

available or default activity ratios to determine the isotopic components of the total alpha results 

before combining the data sets for curve fitting. 

Plutonium at Hanford could have existed as absorption type M or S. Dose reconstructors should 

model 241Am that is a component of plutonium contamination in the lung the same as the plutonium 

matrix in which it has ingrown. In other words, the americium should be treated as the absorption 

type S if the plutonium is type S. 

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 (REDOX) in the early years, PUREX from 1956] and at the Plutonium Reclamation Facility (a 

wing in the Plutonium Finishing Plant). There was also an americium recovery operation in the 

Plutonium Finishing Plant from the late 1960s until 1976. Therefore, waste tanks, transfer lines, and 

parts of the Plutonium Finishing Plant had 241Am that was chemically separate from plutonium. This 

americium should be treated as americium (as opposed to trace americium atoms bound in a 

plutonium matrix) [23]. The International Commission on Radiological Protection (ICRP)-

recommended absorption type for americium is M (ICRP 1994). 

Americium

It has not been discovered yet when americium analyses first started. There is no mention of 

americium excreta analysis in Healy (1948); no mention in a 1954 memo, “Bioassay Annual Report,” 

that lists numbers of urinalyses for plutonium, fission products, and uranium; and no mention in a 

compilation of bioassay procedures, titled “Bioassay Procedures and Analysis (Old Bioassay Bible)” 

(no author or editor), dated April 10, 1961. 

The first americium records in the database show 41 urinalyses for 241Am in 1964 collected from 19 

workers. No americium urinalyses were recorded in 1965 or 1966. The samples were collected over 

a short time period. The first samples, 12 of them, were collected on August 25 or 26. The last 

sample was collected on October 29. The results range from 1.3 × 10-6 to 6.4 × 10-6 [24]. No units 

are given, although the magnitude of the results 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, assuming the units are μCi/L is reasonable [25]. 

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 25/26 samples were baselines 

used 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 [26]. 

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

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

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

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

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

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

database. The duplication appears to be a mistake in the database that does not show up in 

hardcopy listings of these bioassay results. The results are all from special samples for six workers 

who were involved in the same potential intake accident in 222-S on May 2, 1967. There are internal 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 21 of 89

dosimetry evaluations in the workers’ files documenting 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,” by R. C. Henle, dated 

May 1968. The handwritten note says, “5.41 × 10-7 μCi/L detection level.” 

A memorandum to file from John J. Jech (1969a), Senior Development Engineer in the Personnel 

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

new detection limit for 241Am is 2.0 dpm/sample as of July 10, 1969. Mr. 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 Matt Lardy to Harold 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 

(Battelle 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 were 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 described in the Health and Safety 

Laboratory 300 manual (HASL 1985 or similar year) 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 (Battelle 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; however, the MDA is the same (Bihl 

2003). 

Table 5-7 summarizes what has been uncovered concerning 241Am MDAs for routine urinalysis. 

Table 5-7. Routine 241Am urinalysis detection levels [28]. 

Period

MDA, 

dpm/sample

Decision level, 

dpm/sample

1964 4.6E-6 μCi/L Anything detected

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

7/1969 to 2/1974 2.0 Anything detected

3/1974 to 10/1983 0.2 Anything detected

10/1983 to 9/1985 0.04 Anything detected

10/1985 to 05/1988 0.02 Anything detected

05/1988 to 06/1990 0.02 0.01

07/1990 to 10/1991 0.03 0.015

11/1991 to 4/2000 0.02 0.01

5/2000 to 8/2001 0.02 Xb + 2.05 × TPUb

9/2001 to 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.
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 22 of 89

The MDAs listed in Table 5-7 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 5-8. 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 prior to 1969, involving unusual circumstances such as 

using a supposedly sealed source that had ruptured. These intakes were analyzed by urinalysis, so 

obviously a procedure existed at that time, although not part of the contract with UST [29]. On rare 

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

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

level was about 5 dpm/sample (Battelle Northwest 1987). In general, the LEPD result was used just 

as a rapid indicator, and a more accurate result was obtained by wet chemistry/alpha spectrometry 

days later [30]. 

Table 5-8. MDAs for nonroutine 241Am excreta analyses [31]. 

Period

Fecal samples, MDA, dpm/sample Urine samples, MDA, dpm/sample

Emergencya Expedite Priority Emergencya Expedite

1/1967 to 2/1974 (b) NA (b) (b) NA

2/1974 to 1981 (c) NA 4 (c) NA

1982 to 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 to 9/1985 200 NA 0.16 1.0 NA

10/1985 to 6/1989 20 6 0.1 1 0.08

7/1989 to 10/1991e 20 4 NA 2 0.4

11/1991 to 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-RL 1978)] 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 (Larson 

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, probably about 0.2-0.5 

dpm/sample considering the state-of-the-art of those laboratories. 

5.2.3

The curium isotopes of concern were 242 and 244. Most sources of curium at Hanford were minor 

calibration sources or minor constituents in an actinide mixture; however, extraction of 244Cm from 

high-level waste occurred at the 325 Building sometime in the 1970s. 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. After 1985, the 

chemistry is the same as americium, but 241Am, 242Cm, and 244Cm were reported separately if 

requested (Battelle Northwest 1987). The MDAs were not always identical with 241Am, however. 

Routine urinalysis MDAs for curium are provided in Table 5-9 and nonroutine excreta analyses are 

provided in Table 5-10.

Curium 

5.2.4

The history of tritium urinalysis at Hanford is not well documented. Tritium urinalysis was not 

mentioned at all in Wilson (1987). The earliest report found to date on tritium urinalysis at Hanford 

dates to 1949 by Jack 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 

Tritium
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 23 of 89

Table 5-9. Routine Cm urinalysis detection levels [32]. 

Period MDA, dpm/sample Decision level, dpm/sample

7/1969 to 1981 Not specifically mentioned 

1982 to 9/1983 Listed for emergency processing only

10/1983 to 4/1988 0.02 Anything detected

5/1988 to 6/1990 0.02 0.01

6/1990 to 10/1991 0.03 0.015

11/1991 to 4/2000 0.02 0.01

5/2000 to 8/2001 0.02 Xb + 2.05 × TPUa

9/2001 to 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 Cm244 respectively. However, it is recommended that dose reconstructors use 

2 × TPU as the actual decision level for both periods. [33] 

Table 5-10. MDAs for nonroutine Cm excreta analyses [34]. 

Period

Fecal samples, MDA, dpm/sample Urine samples, MDA, dpm/sample

Emergencya Expedite Priority Emergencya Expedite

Prior to 1982 (b) NA (b) (b) NA

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

10/1983 to 9/1985 240 NA 0.8 10 NA

10/1985 to 6/1989 240 70 0.8 1 1.2

7/1989 to 10/1991e NA NA NA NA NA

11/1991 to 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 (Larson 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.

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

g. Probably available but MDAs not found. 

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

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

large samples.

j. 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.

was mentioned in that letter, but one was mentioned in an internal memo (Parker 1950) that referred 

to the acetylene method for urinalysis and 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 described is hard to follow, but it

definitely was not liquid scintillation counting. A 1961 report entitled 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. Liquid scintillation counting was implemented for tritium bioassay at the 

Savannah River Site in 1958 and it is reasonable to expect that Hanford did so at about the same 

time. In the above-mentioned interview with Matt M. Lardy (Bihl 2003), Mr. 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 [35]. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 24 of 89

Tritium was referred to as P-10 in the 1950s (for example, Parker 1950). The main source of tritium in 

the 1950s was 108-B, also called the P-10 Plant, in which tritium was extracted from irradiated Li-Al 

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). 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), and there has been low-level use of tritium as a tracer in various biology experiments. 

Tritium in the groundwater was a source of ingestion for 400 Area workers, principally the Fast Flux 

Test Facility, from about 1978 to 2003. See Section 5.6 for 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. 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, indicating 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, with the smallest value being 

2.5 μCi/L. None of the values are listed as less-than values. The 1965 statement of work with UST 

(AEC-RL 1964) shows an MDA of 1 μCi/L (which is consistent with the MDA at Savannah River Site 

throughout the 1950s). Table 5-11 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 5-11. Routine tritium urinalysis 

detection levels [36]. 

Period MDA

1949 through 1960a ≈ 5 μCi/L

1961 through 1981 1 μCi/L

1982 through 10/1991 10 dpm/ml

11/1991 to present 20 dpm/ml

a. Dates and MDA are best guesses. The change in 

1961 was based on earliest reference to liquid 

scintillation counting.

5.2.5

Uranium exposure at Hanford involved principally three physical forms: depleted (DU), natural (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 may 

have experimented with more enriched uranium at different times (e.g. metallurgy on commercialgrade fuel), but such exposure would have been to small groups for limited periods. For instance, 

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

for the 306 Building effective for November 9, 1970, through November 9, 1971 (Battelle Northwest 

1970; Jech 1970a). Table 5-12 provides the default uranium mixtures that have been used at the 

Hanford Site since the 1980s (Carbaugh 2003a). However, different batches of uranium can have 

different isotopic mixtures; therefore, the default mixtures in the Integrated Modules for Bioassay 

Analysis (IMBA) codes 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, waste management) were exposed to natural uranium during operation of the 

early reactors (through 1951) and recycled uranium starting in February 1952 at the UO3 Plant and 

July 1952 at the 300 Area Fabrication plants (DOE 2002). Recycled uranium had impurities build up 

Uranium 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 25 of 89

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 5-13 (DOE 

2000). Plutonium-239 can be assumed for the plutonium alpha impurity and 232Th can be assumed 

for the thorium [37]. 

Table 5-12. Radiological characteristics of Hanford uranium 

mixtures. 

Uranium mixturea,b

Weight percentage Recycled (RU) Commercial fuel (CF)

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. CF 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.

Table 5-13. Impurities in recycled uranium 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 uCi 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, Bihl, Carbaugh 1989).

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 in 1988 and take into account other information presented in the DOE investigation 

of recycled uranium 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. [38]
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 26 of 89

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 5-14. Because the relationship 

between the old lung fluid studies and the ICRP Publication 66 absorption types (ICRP 1995) is not 

established, Table 5-14 also shows recommended absorption types for intakes from the listed 

facilities, which dose reconstructors should use unless person-specific data are available. These 

absorption type assumptions should be applied to the impurities as well. 

Table 5-14. Inhalation class for Hanford uranium compounds. 

ICRP 30 inhalation class 

from lung fluid studies Compound and location

Recommended ICRP 66 

lung absorption type 

80% D

20% W

Hanford UO3 Plant smear sample dissolution study in 1984a

(UO3 powder)

(b) 

10% D

90% Y

Hanford 303-M Building air sample dissolution studyc (300 

Area Uranium Fuel Production Facilities)

(b)

29% D

71% Y

Hanford 333 Building air sample dissolution studyc (300 Area 

Uranium Fuel Production Facilities)

(b) 

20% D

80% Y

Hanford 306-W Building Machine Shop air sample dissolution 

studyc (b)

Uranyl nitrate at PUREX or UO3 Plant F

UCl4 Md

U carbonate (assumed form after discharge to the soil) Fd,e

a. Sula, Bihl, and Carbaugh (1989).

b. Because the conversions from the solubility studies to the ICRP (1995) 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.

c. Fisher (1986).

d. Long (1993) 

e. Cooke and Holt (1974). [39]

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 used was to obtain both a Friday evening 

sample and Monday morning sample. The period of this sampling scheme was not established, other 

than in the 1970s and maybe earlier. This scheme was changed to Monday-morning-only sampling 

about the early 1980s. Changeover should be clear in the records. 

The Friday/Monday sampling scheme was also used in 1962 and 1963 for 313 and 314 Building 

workers. 

Wilson (1987) states that the uranium urinalysis program prior to 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: Mr. 

Lardy said “about 1990” but other evidence indicates late 1991.) Lardy (1970) describes two 

procedures: one with 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; the other with 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 using 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 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 27 of 89

extraction with hexone (Carbaugh 1989). A 100-ml aliquot was used, but the results were reported as 

per total sample. The chemistry for the KPA involves a 50-ml aliquot that is wet-ashed with acid, 

passed through an ion exchange column, then eluted with weak acid. Results are reported as per 

total sample. 

When alpha spectrometry was introduced in 1983, two uranium urinalyses procedures were offered: 

the elemental procedure discussed above and the alpha spectrometric procedure to provide isotopic 

results. In general, the elemental procedure was used for workers 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 working 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 [40]. 

Alpha spectrometry cannot differentiate between 233U and 234U. Prior to 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 233

U 

was rare after the early 1970s, long before alpha spectrometry came into use for bioassay. 

Therefore, unless specifically mentioned in an intake investigation report, assume 233U results since 

1983 are actually 234U [41].

Table 5-15 summarizes the routine urinalysis detection levels and Table 5-16 summarizes nonroutine 

detection levels. 

Table 5-15. Routine uranium urinalysis detection levels. 

Period

Elemental Isotopic

MDA, 

μg/L

Decision level, 

μg/L 

MDA, 

dpm/sample

Decision level, 

dpm/sample

Prior to 1948 Not specifically mentioned NA NA

1948 through 1949 10 Anything 

detected

NA NA

1950 through 1974a 4 (reporting level) (b) NA NA

1974 through 1981 0.4 Same as MDA NA NA

1982 through 9/1983 0.05 – 0.25 (0.1 

most probable)c

Same as MDA NA NA

10/1983 through 

12/1983

0.03 μg/sample 0.5 μg/sampled 0.035 Same as MDA

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

9/1985 through 6/1990 0.03/0.5 μg/same 0.2 μg/samplef 0.02 Same as MDA

6/1990 through 

10/1991

0.2/0.5 μg/same 0.2f 0.03 0.15/0.015g

11/1991 through 

present

0.06/0.5 μg/same 0.2f 0.02 0.15/0.010g

a. Wilson 1958. Also Lardy (1970) 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 1958-1966, values were reported well below the 4 μg/L value but it is not known why (i.e., was the MDA 

actually much lower or was the laboratory just reporting all values as measured without truncation). 

c. MDAs were based on sample size, but 0.1 μg/L applied to most sample sizes (Larson 1982). 

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

e. 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 [42].

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

g. First value applied to 234U and 238U; second value applied to 235U based on natural background in urine. In 

2002, the 235U decision level was lowered to 0.007 dpm (Carbaugh 2003b). 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 28 of 89

Table 5-16. MDAs for nonroutine uranium excreta analyses [43]. 

Period

Analysis 

type

Fecal samples, MDA, dpm/sample Urine samples, MDA, dpm/sample

Emergencya Expedite Priority Emergencya Expedite

1980b Elemental NA NA NA 10 μg/L NA

10/1983 to 9/1985 Elemental 8 μg/sample NA 0.3 μg/sam 7 μg/sam NA

Isotopicc 12 dpm/sam NA 0.3 dpm/sam 1 dpm/sam NA

10/1985 to present Elemental 8 μg/sam 5 μg/sam 0.3 μg/sam 7 μg/sam 0.5 μg/sam

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

6/1990 to 10/1991d Elemental 20 μg/sam 4 μg/sam NA 20 μg/sam 4 μg/sam

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

11/1991 to present Elemental 8 μg/sam 5 μg/sam 0.3 μg/sam 7 μg/sam 0.5 μg/sam

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

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, probably about 0.2-0.5 

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 recycled uranium through measurement of 236U. The 

presence of 236U confirms an occupational intake of recycled uranium; the detection limit for 236U is 

such that urinary excretion of uranium greater than 0.2 μg/L (see discussion of natural background 

excretion below) from an intake of recycled uranium should have a detectable amount of 236U 

(Carbaugh 2003a, MacLellan 1995). 

Natural uranium 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 depleted uranium from natural uranium but, 

considering uncertainties in analytical results, that ratio cannot be used to distinguish recycled 

uranium. Three studies were conducted, in 1985, 1990, and 1995, to establish the range of natural 

background excretion in unexposed persons living near the Hanford Site. The third study purposely 

looked for possible geographic and seasonal differences in the background. All studies found natural 

excretion to be lognormally distributed. Although the 50th percentiles and slopes of the excretion 

curves were different in the studies, each study found 0.2 μg/d to be about 99 to 99.9 percentile. The 

1995 study had one result that greatly exceeded the 0.2 μg/d value (Carbaugh 2003a). 

Therefore, 0.2 μg/d was established in 1985 and continues to be used at present as the 

environmental decision level for exposures to natural or recycled uranium. 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). 

It is reasonable to use the urinary excretion values of 0.2 μg/d for elemental analyses, 0.15 dpm/d for 

234U and 238U and essentially anything detected for 235U, to distinguish between natural background 

and potential occupational exposure for natural and recycled uranium, unless the worker’s file shows 

236U results or other studies show the excretion was from natural sources. These environmental 

decision levels should apply to the entire history of Hanford. Prior to 1985, there will undoubtedly be 

excretion values exceeding the environmental screening levels that were nevertheless due to natural 

sources, but it is unlikely there will be data available to prove it. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 29 of 89

Background excretion of uranium in feces probably varies over an even larger range than urinary 

excretion; however, 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 

production runs in 1966 and 1970. A total of 820 kg of 233U were produced (Walser 1978, Isochem 

Inc. 1967). See Section 5.3.4 for 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. 

Conversion to U308 at Z Plant was discussed in the planning stage but whether that was actually done 

has not been verified yet (Tomlinson 1964). 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 or the units of the recorded results. Copies of an earlier database show the 

units to be μg/L, consistent with the units for all uranium analyses at that time; however, the recorded 

values are much too small to realistically be in μg/L, with the exception of the single value in 1969. 

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 bioassay because of the high 

specific activity of the 233U and the major contaminant 232U [44]. 

There were contaminants in the 233U. Specifications for contaminants in the 233U product were more 

lax than for RU. The 232U contaminant specification was stated as 6 ppm in the Tomlinson (1964) 

planning document but was listed as 8 ppm in the specifications for the first campaign (Isochem, Inc. 

1967). However, the 232U component would have been accounted for in the bioassay if the bioassay 

was in terms of either micrograms or total alpha activity (less than 2% of the total uranium activity 

would have come from the 232U). The 228Th (half-life 1.9 yr) progeny of 232U would have built up slowly 

in the final product, but this would have occurred after the product was packaged so it would not have 

been a significant source of intake during processing or packaging. Other contaminant specifications 

were also stated in the Tomlinson planning document and updated by the Isochem document. If an 

intake of 233U is assigned for an Energy Employee, add intakes of 62 pCi 239Pu, 2 pCi 95ZrNb, 4 pCi 

106Ru, 1 pCi 233Pa, and 0.21 pCi 237Np per microgram of 233U (6.4 × 10-3 pCi 239Pu, 2.1 × 10-4 pCi 

95ZrNb, 4.1 x10-4 pCi 106Ru, 1.0 x10-4 pCi 233Pa, and 2.2 × 10-5 pCi 237Np per pCi 233U). [Note: the 

106Ru actuals exceeded the specification in about 15% of the batches as reported in the Isochem 

document hence the highest actual was used.] 

See Section 5.6 for directions on how to apply intakes of 233U for unmonitored workers, bearing in 

mind that only a few workers were monitored specifically for 233U and only in 1970. [Workers may 

have been monitored via normal uranium urinalyses, but that method would not have had sufficient 

sensitivity for intakes of pure 233U.]

5.2.6

Fission product urinalysis was the method used to monitor 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 

prior to 1948 should be considered unreliable and ignored. The procedure 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 

Fission Product Analysis
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 30 of 89

in use in 1954 with the addition of a cerium carrier. It was also listed in the compilation of procedures 

referred to as the “Old Bioassay Bible” in 1961, but that same compilation had a separate procedure 

for 90Sr in urine. A memo 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. “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 5-17 summarizes the most 

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

Table 5-17. Routine fission product urinalysis 

detection levels [45]. 

Period MDA 

1948 to 2/1956 60 dpm/samplea

3/1956 to 10/1964 70 dpm/sampleb

1965 to approx. 1969 31 dpm/sample

1970c 5 dpm/sample

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 products and activation products to which a worker could have been 

exposed. The procedure separated and counted radionuclides of alkaline earths 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. OTIB-0054, 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. See also Section 5.4 for 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 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 [46]. 

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, 1961, had a procedure specific 

for strontium in urine and fecal salts that included counting total strontium and then allowing for 90Y 

ingrowth, yttrium separation, and counting of 90Y to account for 90Sr separate from gross strontium 

beta, if desired. This procedure was also mentioned in a memorandum, dated July 1963, 

documenting discussions between the Analytical Laboratories and Internal Dosimetry clarifying 

logistics of handling these samples and reporting 90Sr results. Handwritten notes on this memo 

Strontium
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indicate that the detection level is about 20 dpm. Nevertheless, the database records show the use of 

fission product urinalyses into 1964 and 90Sr urinalyses apparently starting in 1965. The database 

contains frequent entry values 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), dated August 1964, that listed a detection limit for 90Sr as 25 pCi/1.5L, which 

converts to 56 dpm/1.5L or 37 dpm/L (AEC-RL 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 discussing 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-RL 

1978) except the analytical limit is defined as ±100%. A letter (Robinson 1979) 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 (Battelle Northwest 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 (Battelle 

Northwest 1992). However, the procedure stayed the same throughout this period and the true MDA 

probably held at about 2 dpm/sample.

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

truncated at a reporting level. 

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 [47]. 

When this happens, the result is reported as Sr total or SRTOT, but the result may be interpreted as 

90Sr. These results were below the required detection level anyway. Table 5-18 summarizes the 

routine urinalysis detection levels for 90Sr procedure.

Table 5-18. Routine 90Sr urinalysis detection levels [48]. 

Period MDA or MDC

Prior to 1965 Might have been available but MDA not known

1965 to 1969 1.67E-5 μCi/L (37 dpm/L)

1970 to 1974 1E-6 μCi/L (2.2 dpm/L)

1975 to 3/1979 50 dpm/samplea

4/1979 to 1981 5 dpm/sample

1982 to 9/1983 2.5 dpm/sample

10/1983 to 6/1990 2 dpm/sample

9/1990 to 11/1991 30 dpm/sample

11/1991 to 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. Prior to that time, the MDA was also 

used as the decision level. 

Dose reconstructors should consider all strontium at Hanford to be absorption type F [49]. It is 

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

might be present.
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5.2.8

Hanford was involved in the manufacture of heat sources using 147Pm. The period seems to start in 

1966 and continue into the early 1970s (Howell and King 1968). The high activity work (kilocuries) 

took place in the 325 Building, but some exposure apparently occurred as early as 1962 or 1963 in 

the 222-S Chemistry Laboratory [50] and the 325 Building and as late as 1971 in the 308 Fuels 

Laboratory (Howell, 1969; Glover 1964). 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 human volunteers study (14 

volunteers) using 143Pm was conducted in 1967 or 1968 (Palmer, Crook, and Nelson 1969, 1970) (12 

volunteers: injection of 0.1 μCi each; 2 volunteers: ingestion of 10 μCi each). Another study on the 

behavior of promethium used swine with 148mPm as the surrogate for 147Pm (McConnon et al. 

undated). 

Promethium

The work on the heat sources involved converting promethium/cerium nitrates into Pm2O3 by 

separation chemistry then calcining (Howell and King 1968). There was also one mention of coldpressed, 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 μCi/L for urine and μCi/kg for feces through 1974 and dpm/sample from 

1975 to the present. Table 5-19 lists the 147Pm minimum detection levels at various times.

Table 5-19. Routine 147Pm urinalysis detection levels [51]. 

Period MDA or MDC

Prior to 1965 Might have been available but MDA not known

1965 to 3/25/1970 1.67E-5 μCi/L (37 dpm/L)

3/31/1970 to 1974 1.0E-5 μCi/L (22 dpm/L)

1975 to 1979 50 dpm/samplea

1980 to 1981 20 dpm/sample

1982 to 9/1983 5 dpm/sample

10/1983 to 6/1990 4.0 dpm/sample

11/1991 to presentb 30 dpm/sample

a. 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. 

b. No 147Pm analyses were performed during 1990 and 1991. 

Fecal samples were analyzed for 147Pm for some of the potential intake events in the late 1960s. The 

MDA or at least the lowest reporting level appears to be 1.67 × 10-5 μCi/kg [52]. An MDA for fecal 

samples does not appear in laboratory statements of work during the 1970s, but reappears in the 

1980s: 28–110 dpm/sample in 1982 depending on sample size (roughly 400 dpm/kg); 

220 dpm/sample from 1983 to the 1990s [53].

Only one description of the procedure was found, and that procedure appeared in documents dated 

1970, 1974 and 1977 (Lardy 1970, UST 1974, 1977). 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 pH less than 1, then at pH 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. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 33 of 89

5.2.9

Considerable activity toward initiating a bioassay procedure and establishing a biokinetic model for 

210Po was found in the files from about 1968 through the mid-1970s. There is an indication of work 

with pure 210Po in the 308 Building in 1968 and again in 1975. Whether the work in the 308 Building 

was continuous through that period or just in those two years was not determined. Inference can be 

made that there was work somewhat prior to 1968 based on a handwritten note documenting a 

telephone conversation in November 1967 during which it was stated that the 210Po starts in the 

process in the soluble form but is converted to the insoluble form (Henle undated). However, UST 

was asked to develop a bioassay procedure in March 1968 and did so shortly thereafter, so 

apparently concern for possible intakes became important in early 1968. In addition, work with 210Po 

in the 325 Building started in June 1972 and was scheduled “to run for 3.5 years” (Jech 1972). 

Polonium

The procedure developed for 210Po by UST 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 hr. 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 5-20 and for nonroutine excreta bioassay in Table 5-21. 

Table 5-20. Routine 210Po urinalysis detection levels [54]. 

Period MDA or MDC

3/1968 to 1970 5.4E-7 μCi/L

1971 to 1974 5.0E-7 μCi/L

1975 to 1979 1 dpm/samplea

1980 to 9/1983 0.1 dpm/samplea

10/1983 to 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 5-21. MDAs for nonroutine 210Po excreta analyses [55]. 

Period

Fecal samples, MDA, dpm/sample Urine samples, MDA, dpm/sample

Emergencya Expedite Priority Emergencya Expedite

3/1968 to 1973 NA NA 5.4E-7 μCi/kg NA NA

1974 to 9/1983 NA NA (b) NA NA

10/1983 to 9/1985 340 NA NA 0.8 NA

10/1985 to 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, 210Po exists naturally in urine and 

feces. Nothing was found in the records indicating that a study on natural excretion levels for persons 

living around Hanford had been conducted. 

Bismuth slugs were irradiated in Hanford reactors to produce 210Po starting in 1945 and continuing 

into the early 1950s; however, it appears that the slugs were promptly shipped offsite (Tilley 1945, 

DOE 1997). No evidence has been found that extraction or handling of concentrated 210Po was 

conducted at Hanford during the early years. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 34 of 89

5.2.10

At PUREX from 1958 through 1972, 237Np was removed from the dissolved fuel, purified, and 

packaged for shipment offsite. It was downloaded from an ion exchange column and packaged in 

liquid form, but the chemical form has not been discovered yet. Although mostly 237Np by mass, the 

small mass of 238Pu in the product produced most of the radioactivity. Plutonium bioassay was 

considered sufficient to monitor for intakes [56]. 

Neptunium 

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 μCi, but it has not been determined to date if the units are μCi/sample 

or μCi/L. 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 dpm/sample [58]. The 1983 statement of work with UST (Battelle Northwest 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 Pacific Northwest National Laboratory. As such, small-scale (in terms of either the number 

of persons or activity of the source) use of various radionuclides not addressed above has occurred 

throughout the history of the Hanford Site. The following discussion, addressing 14C, 232Th, radon, 

90Y, 227Th, 227Ac, and 32P, is probably not comprehensive.

Other Limited-Exposure Radionuclides

Carbon-14 exposure occurred at the 3731 Building in the mid-1950s when irradiated graphite 

samples were brought to the building from the operating reactors for destructive testing. No 

information has been uncovered about what bioassay, if any, was done. Carbon-14 was also used as 

a tracer in biological experiments. One documented study was conducted in the late 1990s in the Life 

Science Laboratory-II Building, involving a total of about 4 Ci of 14C. Urinalyses were obtained on 

about 20 researchers. The MDA was 10 dpm/ml (Battelle Northwest 1987, Lyon et al 1991, Long 

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. 

There are some thorium urinalysis samples 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 using alpha spectrometry. The detection level was 

listed as 0.5 dpm/sample in 1970 (Lardy) and 0.03 dpm/sample in 1983 (Battelle Northwest). Some 

work was also done with 232Th slurries in the 3720 Building in the mid-1990s. The plan was to collect 

baseline urine samples on the few workers involved, 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). 

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 μCi/L and the reporting level was apparently 1 μCi/L.
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A radon generator was used for animal studies in the 108F Building; it was later moved to LSLII. 

Monitoring was probably just by air sampling; but no information has been discovered yet. There 

should have been only a few researchers potentially exposed [59]. 

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-hr 

physical half-life; however, the need to perform chest counting never arose (Carbaugh 1990, Bauman 

1996). 

Another project 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; however, the project ended after only a few milkings. Only a couple of 

researchers were involved (Bihl 1995). The bioassay procedure had a stated MDA of 0.1 dpm/sample 

for 227Th. 

Phosphorous-32 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. 

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-ray and gamma-ray-emitting radionuclides since 1960. (Unless 

otherwise noted, the in vivo information below came from Wilson 1987 and Lynch 2001). The unit for 

all in vivo measurements is nCi.

5.3.1

The first whole-body counter started counting workers in mid-1959 and became a routine method in 

1960. It consisted of a single NaI crystal (9.375-in. diameter and 4-in. thick) housed 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, copper) (Wilson 1987; Roesch, McCall, and Palmer 1960). This was called 

the Iron Room. The counting geometry was a chair configured to simulate a 1-m arc. The original 

count time was 20 min, which was reduced to 10 min in October 1962 (Swanberg 1962). A second, 

same-sized 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 

quoted in a report in the fall of 1964 were the same MDAs listed in Mr. Palmer’s Laboratory Record 

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

warrant republishing the MDAs. Therefore, the MDAs shown in Table 5-22 are the only MDAs found 

for the 1960s and 1970s, and apparently were meant to apply generally to the various whole-body 

counters in operation during this period. Although they were referred to as detection limits, the 

calculations were more consistent with what is presently referred to as decision levels. 

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 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 36 of 89

Table 5-22. Routine whole-body counting detection levels.

a

Period Nuclide MDA (nCi) Decision level (nCi) Reporting level (nCi)

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 electronic 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
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 37 of 89

Period Nuclide MDA (nCi) Decision level (nCi) Reporting level (nCi)

10/1999 to 

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 HPGe system 

(Lynch 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). 

and sides by lead. The bed moved under a large NaI crystal (11.5-in. diameter by 4-in. thick) that was 

also shielded by lead with the exception of the downward-looking face that looked directly onto the 

body as it passed under the crystal. The shadow shield detector was mounted in a mobile trailer and 

moved to areas nearer the Hanford worksites. The trailer also had a thyroid detector and a wound 

counter. The shadow shield detector became operational in 1963 (Brady 1964). The mobile counter 

was described as having comparable sensitivity to the “larger, conventional whole body counters 

installed in massive iron rooms. There is, however, some decreased sensitivity in the lower energy 

region below about 300 keV, due to increased contribution to the background from scattered 

radiation” (Swanberg 1963).

A report on the radionuclides detected in workers at the whole-body counter facility in 1961 listed 

24Na, 60Co, 65Zn, 95Zr, 95Nb, 99Mo, 99Tc [presumably 99Mo], 103Ru, 106Ru, 131I, 137Cs, and 144Ce (Henle 

1962). A similar report summarizing 1961–1963 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% 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 listing 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 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/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 located behind the thoracic region. Count time is 200 sec. (Palmer et al. circa 

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 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 38 of 89

bed. The configuration of this system, in terms of number and size of the detectors, has changed 

many times. It started as four 68% HPGe detectors; one of the detectors was replaced with a 120% 

detector in late 1995; in May 1997 the system was upgraded to include seven detectors including 

three 120% detectors. As a four-detector array, the system was used only when 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 being used for routine counts for workers exposed to mixtures of 137Cs and plutonium. The 

count time was usually 10 min; however, 20-min count times are used as confirmation of an initial 

count with detectable activity. As a consequence, the database will usually show a 10-min count and 

a 20-min count on the same day or a few days later if the first count had detectable activity (excluding 

40K or medical radionuclides) (Lyon et al 1998). 

The radon progeny 214Bi and the thoron progeny 208Tl were and are present in the background spectra 

from the shielded rooms used for in vivo measurements of workers at Hanford. To account for this 

contribution when calculating worker results, the background levels of these nuclides were estimated 

from background measurements in the rooms with the detectors shielded with an appropriate 

phantom. The environmental background count rates in addition to the continuum background count 

rates were 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 using the coaxial germanium counting system. The DIL values were 

6.47 nCi for 214Bi and 0.75 nCi for 208Tl and represented the upper 99% confidence level of 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 surrounding 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 

tying the intake to an acute event, the dose reconstructor can assume either chronic ingestion or 

inhalation according to which is more favorable to the claimant for the organ of concern. Chronic 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 39 of 89

ingestion intakes would have been occurring prior to implementation of whole-body counting, probably 

from shortly after the startup of the first once-through-cooling reactor (August 1944), and would have 

ceased shortly after the shutdown of the 100 KE reactor (January 1971) [66]. 

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 and 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 hardcopy 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 used 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 count. 

• 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 1987). 

Table 5-23 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 5-23, the dose 

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

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 min. 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 min. (Wilson 1987). A lung phantom with variable chest wall thickness 

was

Chest Counters
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 40 of 89

Table 5-23. Mean body burdens of 137Cs from fallout in the 

United States.a

Year Body burden (nCi) Year Body burden (nCi)

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 (1987).

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, depending on 

a subject’s weight-to-height ratio (Andersen 1971, Wilson 1987). (Chest count MDAs are summarized 

in Table 5-24.) 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]. 

A solid state germanium counting system using three planar HPGe detectors (shortly upgraded to six 

detectors) replaced the NaI detector in the Iron Room chair counter in 1983 (Palmer et al. 1984). The 

HPGe detectors provide 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 faces the chest for better transmission of lowenergy 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 sec. 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 RDA or Reliably Detectable 

Activity, 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 sec with nearly the same RDAs (Palmer et al. 

circa 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 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 41 of 89

Table 5-24. Routine chest counting detection 

levels. (Andersen 1971, Lyon et al. 1988, Lyon et 

al. 1989, Lyon et al. 1990, Lyon et al. 1993, Lyon 

et al. 1994, MacLellan et al. 1999, Lynch et al. 

2000)

Period Radionuclide MDA (nCi)

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 0.5-0.15a

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 

95th percentile male chest wall (.2/.13) (Lynch, 2003)

c. Adjusted for 95th percentile male chest wall. 

d. Cerium-144 and 154Eu no longer automatically reported 

for chest counts because they now can be quantified in 

the Ge 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. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 42 of 89

increased to 2,000 sec for the six-detector array and 3,000 sec 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 sec for the larger area arrays in November 1995; special counts 

and recounts were 3,600 sec. 

Ultrasonic measurements of chest wall thickness for workers who had activity in the lung began in 

about 1978 and continue 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]. 

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 recently separated from dissolved fuel was certainly 

not in equilibrium, and uranium being treated at the UO3 Plant might have been in equilibrium 

depending on how long it had taken the material to 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, uranium metallurgy studies) [72]. 

5.3.3

Note: During some years, 131I measured in workers who had radioiodine medical studies or treatment 

was recorded in the database; the workers’ files should have a hard copy in vivo count form noting 

that 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 any impact on the 

probability of causation.

Thyroid Counters 

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

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

“At the present time routine thyroid monitoring is conducted on a limited basis in the REDOX and 

PUREX facilities. Generally the pattern for coverage in the PUREX facility includes about four to five 

employees weekly picked from the sampling crews, crane operators, and a Radiation Monitor 

assigned to the stack area. At the REDOX facility routine monitoring is accomplished on a weekly 

basis for the shift crane operators.” (Wilson, 1960) The letter continues to discuss counts and other 

data obtained in 1959; however, 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.” 

The first mobile whole-body counter had a thyroid counter consisting of a 3-in. 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-min count (Wilson 1987). The same detector and MDA were included in a 

description of in vivo counting capabilities at the Whole Body Counting Facility (Andersen 1971) and 

again in 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. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 43 of 89

There is a cluster of thyroid counts with MDAs ranging from 0.020 to 0.050 nCi and another cluster 

ranging 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, not a thyroid calibration [74]. 

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-min count or 0.07 nCi for a 10-min count, but there was no 

mention of which count time was regularly used. There probably were not many workers exposed to 

125I on a regular basis; however, there are indications of a contamination spread in 1978 involving 

several workers (Palmer 1979). The same counter is described for thyroid counting in 1982, except 

the “reporting level” is given as 0.020 nCi; it is not known if this better sensitivity came from a longer 

count time, better positioning, or an improved crystal. 

By 1985, thyroid counting for 125I was performed using two intrinsic germanium detectors, with an 

MDA of 0.005 nCi for a 2,000-sec count. 

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

5.3.4

(Note: Information on use of thorium at Hanford came from Gerber 1992 unless otherwise indicated.)

Thorium Exposure and Monitoring

Thorium contamination was introduced into the 313 Building and surrounding fuel warehouses in the 

early 1950s, when a number of attempts were made to fabricate metallic thorium fuel targets to 

produce 233U. Many problems connected with the rapid formation of a thick coat of oxide on the 

thorium metal targets led to experiments with a variety of bonding methods and coatings. After a 

technique was determined, powdered thorium oxide fuel targets (wafers) were fabricated in the 3732 

Building from 1965 to 1967. Sintering of the wafers was part of the process in the 3732 Building. A 

new technique involving pelletized targets replaced the 3732 Building operations. The pelletized 

targets were fabricated in the 3722 Facility from 1968 through 1970. In addition, that building housed 

a furnace for the “recycling” (reduction) of depleted thorium oxide after it was processed in PUREX. 

Other buildings with thorium contamination included 3307 A and B (change houses), 306 (alloy and 

fabrication test and development activities), and 3706 (radiochemistry in support of fuel fabrication). It 

is reasonable to assume possible exposure to thorium in these facilities as shown in Table 5-25. It is 

reasonable to assume significant exposure, similar to exposure to uranium at Hanford, in the 3732 

Building from 1965 through 1967 and the 3722 Building from 1968 through 1970. If building 

information is not available, but there is evidence that a person worked with uranium in the 300 Area, 

1950-70, it is favorable to claimants to assume exposure to thorium. 

Table 5-25. Buildings with thorium processing

(Gerber 1992, Isochem 1967, Walser 1978).

Building, area Date

313, 300 1950–1970

306, 300 1956–1970

3706, 300 1954–1963

3732, 300 1965–1970

3722, 300 1968–1970

202 A, PUREX, 200 East 1965–1970
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 44 of 89

In addition, the thorium was irradiated in a reactor and dissolved in PUREX, the residual thorium and 

freshly made 233U were separated, and the thorium was recovered (Isochem 1967, Walser 1978). 

This work was performed in three distinct campaigns with extensive cleaning of the PUREX lines and 

tanks prior to introduction of the irradiated thorium wafers. The campaigns were a small test 

campaign conducted in 1965 and two production campaigns in 1966 and 1970. For the 1965 

campaign the original dissolution was not conducted at PUREX but additional purification of the nitrate 

solution was (Isochem 1967, Walser 1978). Handling of the 233U product and the recovered thorium 

was apparently done by the same crew, so thorium exposure should be assumed for PUREX workers 

if there is evidence of bioassay for 233U (Jech 1970b). When not recycled to the 3732 Building, the 

thorium nitrate was stored in the WR Vault in the 200 West Area (Isochem 1967).

Thorium exposure would have been associated with uranium exposure. However, thorium was not 

associated with U Plant so uranium workers in that facility should not be assigned thorium intakes. 

(There were a few episodes of thorium handling at UO3 plant in 1967, involving a few 100s of pounds, 

but this is considered to contribute insignificantly to dose considering that the UO3 plant processed 

about 1,500 metric tons of uranium during 1967 [DOE 2000].) 

Radiation protection practices were similar for thorium and uranium, so assume chronic intake. 

Because metal, oxide, nitrate, and perhaps other forms of thorium were being handled, either 

absorption type M or type S can be assumed (independently from uranium absorption types). 

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); 

hence, the thorium was assumed to be 0.5 yr since purification. Therefore, for an intake of 232Th by 

activity, assign 228Ra at 0.058 times the 232Th intake and assign 228

Th at 0.84 times the 232Th intake

[75]. The 230Th specification for the thorium oxide was <1 ppm (Douglas United Nuclear 1965), which 

means that 230Th contamination would have contributed <6% of the total alpha activity and can be 

ignored. 

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

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

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

28 apply to whole body counts. Urinalyses were used for special investigations. Jech (1967) 

stated the detection limit for 232Th was 1.2 nCi. In another letter, Jech (1969b), summarizing results of 

a worker’s whole-body count, 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 electronic 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 5-26 should be 

used to determine the intake of 232Th per nCi of 228Ac (reported as 232Th) measured in a whole body 

count. Once the adjusted intake of 232Th has been determined, the ratios given in the paragraph 

above must be used to determine the intakes of 228Ra and 228Th.
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 45 of 89

The dose reconstructor can use the whole-body count record plus urinalysis records, if there are any, 

to determine intakes of 232Th. If a person worked in the buildings in Table 5-25 for the years listed, 

and was monitored for uranium but not thorium, it is reasonable to assume exposure to thorium and 

Table 5-26. 232Th intakes from whole body count measurements [78]. 

Period of chronic 

intake prior to whole 

body count (days)

Ratio of Th-232 to Ac228 in whole body at 

time of measurement

Intake of Th-232 per 1 nCi of Ac228 (reported as Th-232) measured 

in whole body count (pCi/d)

Type M Type S Type M Type S

182 23.17 11.33 1810 1310

364 26.56 9.218 1150 648

546 29.03 7.834 874 412

728 30.40 6.898 704 297

910 30.73 6.234 580 232

1092 30.28 5.743 483 191

1274 29.34 5.370 405 162

1463 28.09 5.068 341 142

1645 26.79 4.837 291 127

1827 25.49 4.649 251 115

2009 24.24 4.494 218 106

2184 23.10 4.369 192 98.6

use the coworker intakes provided in the most recent version of the coworker TIB (ORAUT 2007). For 

PUREX workers the preceding statement applies to workers with 233U bioassay for 1965 through 

1970. 

Except for small scope, limited time projects that may have occurred occasionally, exposure to 

thorium at Hanford stopped in 1970. But later-day whole body counts might still be used as a check 

on the intakes estimated for exposure prior to 1970. Th-232 was not routinely reported in whole body 

counts, but a prominent 228Ac peak would have been noticed and investigated. Tables 5-27 and 5-28 

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 level used for a radionuclide specifically targeted for identification by the 

software, and the trigger level used to flag 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 burdens less than that 

might have been overlooked [79]. 

Radon-220 (thoron) was driven off during the bake-out and sintering processes (West 1965, Bihl and 

Traub 2006), but would have grown back in quickly. Measurements of airborne thoron concentrations 

were not found. An upper bound estimation of thoron concentrations in the 3722 and 3732 Buildings 

was made based on the throughput of thorium during two campaigns (1964-1965 and 1969) and the 

following assumptions (Bihl and Traub 2006):

• 90% equilibrium between 232Th and 228Th in the material being processed in the 3722 and 

3732 Buildings (upper bound calculation) 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 46 of 89

• 100% release of thoron from the thorium metal or thoria during the processing (due in part to 

vaporization of 224Ra)

Table 5-27. 232Th to 228Ac ratios in the whole body after end of intake 

period, type M [80] 

Years after end of 

chronic intake

Period of chronic intake, yr

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

period, type S [81]. 

Years after end of 

chronic intake

Period of chronic intake, yr

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

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
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 47 of 89

Years after end of 

chronic intake

Period of chronic intake, yr

1 2 3 4 5 6

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

• 10% of the airborne thoron is not exhausted outside by the oven hoods and escapes into the 

workspaces

• 5 building air exchanges/hr 

• An equilibrium factor of 0.04. 

The estimated working level months/yr (WLM/yr) were 43 (Bihl and Traub 2007). 

These WLM are constant upper bounds associated with processing hundreds of tons of thorium in the 

3722 and 3732 Buildings, 1964 through 1970. All workers associated with these buildings (might be 

just mentioned as thorium processing) should be assigned these WLM. The WLM can be prorated for 

partial year exposure. Use the thoron TIB (NIOSH 2006) to convert the WLM to doses to the ET1, 

ET2, and lung. 

Ventilation in the PUREX canyon would have removed thoron to the main stack; work with small 

quantities of thorium in hoods prior to 1964 would not have produced significant thoron intakes by 

workers [82]. 

Although there was residual thorium contamination in the plants after 1970, the exposure to the 

workers in relation to their uranium exposure would have been negligible and can be ignored. 

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 will almost 

always be 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 1979). Since the mid-1990s, 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 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 “trace” was found. 

More information might be available on the In Vivo Counter Results Form if it was sent to the worker’s 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 48 of 89

personal radiation exposure history file. If not, assume the result of the count is 100 nCi (Glenn 

1968). 

Prior to the advent of GeLi detectors, when a significant peak in a whole-body count of a radionuclide 

not 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 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; include the activity of the interfering radionuclide as given as well. Use 

100 nCi for the interfering radionuclide if not given directly (Glenn 1968). 

The radionuclides routinely reported to the database 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; 

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 using 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 1999) [91]. 

Reporting of radionuclides at levels below the MDA or decision level should not be interpreted 

as implying exposure to those radionuclides. 

For chest counting, the database usually lists 234Th as the potentially measured radionuclide as an 

indicator of 238U. Until recently, routinely reported radionuclides for anyone receiving a chest count 

were 241Am, 234Th, and 235U. This does not imply exposure to both plutonium/americium mixtures and 

uranium. Very recently, workers have been scheduled for types of chest counts based on their 

exposure in the workplace, so for plutonium workers, for instance, only 241Am results are determined 

and reported. 

5.4 MIXTURES

Except in a few facilities in the weapons production cycle (such as B Plant/WESF after 1968, 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 radionuclides that were 

most common. Unmeasured radionuclides generally do not have a big impact on dose but might 

target different organs or might have a larger relative impact over times 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 all 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. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 49 of 89

5.4.1

For fission and activation product mixtures up through 1987 (when N Reactor shut down) use the 

intake mixture ratios presented in Fission and Activation Product Assignment for Internal DoseRelated Gross Beta and Gross Gamma Analyses, OTIB-0054. So if an intake of 90Sr, 137Cs or 106Ru 

was being assigned to a worker, the other fission and activation products should be assigned as 

discussed in that document. OTIB-0054 includes guidance on intake mixtures that applies to workers 

in different types of work or facilities. 

Fission and Activation Products

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 

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 OTIB-0054. 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 (FFTF) for 1980 through 1994 [92]. From 1960 to date these 

radionuclides would have been measured by whole body counts. If their activity is not included in the 

whole body count record, use the standard missed dose procedure for assigning these intakes to 

workers at the applicable facilities using MDAs or reporting levels in Table 5-22. 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 predating whole body counting, intakes of these radionuclides should 

be added using guidance in OTIB-0039 (startup to 1959). 154Eu is not addressed in OTIB-0039 nor 

are MDAs available in Table 5-22 for all years; when not listed in Table 5-22, assume the MDA is 3 

times the MDA for 60Co.(1964 through 1984) [93]. 

OTIB-0054 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 137

Cs in the sludge (Carbaugh 1995, 

Boothe 1992). For workers in waste management facilities or 200 Area workers in general if work 

location is not known, if an intake of fission products is determined and the worker does not have 

bioassay for plutonium or americium after the intake date, also assume intake ratios of 0.001 for both 

239Pu and 241Am relative to the highest fission product intake (Carbaugh 1995, Boothe 1992). 

Tritium intakes may have occurred by reactor workers and fuel dissolution plants workers. Dose from 

tritium was accounted for as part of external dose records for 1949 through 1987. See Section 5.6 for 

guidance on tritium 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. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 50 of 89

5.5 INTERFERENCES, UNCERTAINTIES

5.5.1

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

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 will show up 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 obtained the same day or within a few 

days shows a dramatic decrease in activity or no detectable activity, assume external contamination. 

Radon progeny and medical diagnostic or therapeutic procedures involving 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, use the results as recorded.

5.5.2

Uncertainties for the bioassay measurements were included in the database starting in late 1981 for 

excreta measurements [94]. These are listed in the database under Error and represent total 

propagated uncertainty (one sigma) including counting uncertainty, yield uncertainty, and various 

other systematic uncertainties (UST 1987). Use these when available. For excreta, uncertainty can 

also exist in the sample date. For routine samples, 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 2–3 days of a 

known intake. 

Uncertainties

The period the sample represents is also a source of uncertainty. Most urine samples at Hanford 

were 24-hr simulated samples (kit code 1), meaning the sample was collected over two eveningthrough-morning periods. Medley, Kathren, and Miller (1994) indicated that this sampling method 

produced only about half of a true 24-hr 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-hr 

sample [95]. 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-hr excretion. In general, the error 

associated with collection period results from under-collection of a 24-hr volume. It is favorable to 

claimants to normalize a volume that is less than reference man or reference woman; however, 

volumes larger than reference man or reference woman should be considered 24-hr samples without 

normalizing. If no volume information is available for a urine excretion result, a reasonable 

maximizing assumption would be to double the uncertainty of the result [96]. 

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]. Total 

propagated error has been determined and submitted to the records since then. The propagated 

uncertainty 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 

associated with reproducibly positioning a person to get the same result was studied at Hanford and 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 51 of 89

found to be about 5% [99]. All calibrations use phantoms, and there is considerable uncertainty 

associated with 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 1332 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 associated with 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 (Palmer 1990 Table 7-4). Uncertainties would be 

much higher for an individual with activity in the bone and/or liver. The uncertainty in lung activity 

estimates 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/weight correction was at least 50% 

for the front/back lung counter. 

Because the uncertainty in the 241Am-to-Pu-alpha ratio can vary nearly two orders of magnitude, 

plutonium intakes should not be determined solely by chest counting data if possible. If no 

information concerning the isotopic mixture of an intake is available and a default mixture is assumed, 

a geometric standard deviation of 5 uncertainty should be associated with an intake determined by 

chest counting [102]. 

Based on the above discussion, the assumption provided in the Internal Dose Reconstruction 

Implementation Guide (NIOSH 2002), namely the standard deviation is 0.3 times the MDA or 

reporting level, is adequate with the exception of chest counts, for which the dose reconstructor 

should use 0.5 times the MDA. If actual standard deviations or other indications of error are 

reported with a bioassay measurement result that exceeds the above guidance, use the reported 

value. 

5.6 UNMONITORED WORKERS

For unmonitored workers with external dosimetry, use the intakes provided from the coworker 

analyses in OTIB-0039 (ORAUT 2007a). Short-lived radionuclides, such as 144Ce, 106Ru, 95Zr, 95Nb, 

54Mn, 65Zn, 140Ba, 140La, 51Cr, 59Fe, 110mAg, should not be assigned to workers after December 31, 

1988 [103]. The exception is for 24Na for workers at the Fast Flux Text Facility, which operated 

intermittently from 1980 to 1984. Special consideration is also provided for workers involved in 

decontamination and decommissioning. See facility-specific details at the end of this section. The 

coworker intakes do not address tritium or radioiodines. Guidance for assignment of intakes of tritium 

and radioiodines by unmonitored workers is provided in this section. There is also guidance for 

assignment of intakes of 233U and contaminants for unmonitored workers on the 233U project. 

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-level intakes. 

Before 1994, workers with even a remote chance of exposure to workplace external radiation or 

workplace airborne contamination (as opposed to exposure from effluents) wore a dosimeter. 

Through 1971, workers obtained film dosimeters before proceeding past badge houses at the various 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 52 of 89

facilities. From 1972 to 1994, even workers with only a small potential for incurring external dose 

wore a minimal dosimeter, called the Hanford basic dosimeter (ORAUT 2006a). 

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

[104]. Termination whole-body counts were often provided to any worker regardless of exposure risk 

and should not be considered evidence of being on a bioassay program.

Considerable (hundreds per month) thyroid scans were being done for workers in the fuel dissolution 

(canyon) buildings during 1945 through at least 1946. 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. The Ikenberry document 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 considered to be reasonably due to a thyroid burden (as opposed 

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 (peak production years), only about 130 workers had detectable 131I out of 

about 14,500 whole body and thyroid counts; results ranged from 0.001 to 560 nCi, the highest 

measurement being a same-day measurement after an accidental exposure in the T Plant canyon. 

Ninety percent of the detected results were less than 15 nCi [105]. From these two data sets, the 

assumption was made that was favorable to claimants that at-risk workers had thyroid burdens less 

than 30 nCi, which would result from a daily intake of 20,000 pCi [106]. This intake would apply to all 

at risk workers prior to 1960. From 1960 to 1987, it is reasonable to assume that an unmonitored 

worker, (i.e., did not have a whole body count) did not incur intakes greater than the thousands of 

workers given whole body counts who showed no detection. Using a nominal detection limit for whole 

body counts (as opposed to more sensitive thyroid counts) of 5 nCi, unmonitored workers associated 

with reactors or fuel dissolution plants should be assigned 131I intakes of 3400 pCi/d [107]. No intakes 

of 131I should be assigned after 1987 because there was no significant source term, unless, of course, 

the worker’s personal information indicates a special exposure, such as for a research project [108]. 

Tritium bioassay results are not included in the electronic database until 1984 [109]. Tritium doses 

were included in external dose reports from 1949 through 1987, so “monitored” workers would be 

determined by tritium dose listed 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 under “tritium” might be 

indicative of either an unmonitored worker or a monitored worker with 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 [110]. The 108B Tritium Extraction Facility shut down in 1955 so that probably 

explains the lack of tritium doses, and probably a lack of tritium monitoring, until the heavy-watermoderated Plutonium Recycle Test Reactor started up in 1960 (see Sections 2.1.6 and 2.2.8). 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 5-29 should be applied. 

Prior to 1949, it is assumed that tritium intakes were limited by the tolerance level. The earliest tritium 

tolerance limit found to date is an air concentration limit 0.01 μCi/L in 1947 (Parker 1947, Peterson 

1949)(See Appendix 5A.2 for discussion on tolerance levels). Assuming chronic intake at this 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 53 of 89

concentration of 1.2 m3/hr for 2000 hr per year results in an intake rate of 6.6 × 107 pCi/d. This 

produces annual organ doses of about 1,500 mrem [111], which is larger than all but a few recorded 

doses from 1949 through the 1960s. Hence, this intake rate is favorable to claimants for the highest 

risk workers, namely workers at the 108B 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 108B facility. It is assumed that 

tritium intakes at the reactors and fuel dissolution facilities would have been at most 10% of the 

intakes at 108B [112]. Therefore, daily intakes of 7 × 106 pCi/d should be assigned to unmonitored 

workers at the reactors, fuel dissolution plants, or rovers.

Table 5-29. Statistical parameters of Hanford recorded tritium doses and 

associated intakes [113].

Year

Geometric 

mean 

(mrem) GSD

Geometric mean 

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

1955–1960 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 CADW.

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

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, dose 

from tritium <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 and geometric standard 

deviations (GSD) are shown in Table 5-29. 

The methodology for calculating tritium doses from urine concentration has changed over time. The 

method introduced in ICRP Publications 2 and 10 (1959 and 1968) resulted in dose per unit activity in 

urine slightly higher than the method used presently. How the tritium doses were calculated prior to 

1959 has not been found yet; however, Parker, head manager of radiation protection at Hanford in the 

1940s and 1950s, referenced a method developed by Morgan in 1946 which 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. Assuming the tritium is distributed in 42 L of body water, at equilibrium, 952 μCi/L 

[40,000 μCi/42L] produced 100 mrem/d; or, 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 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 54 of 89

permissible dose of 300 mrem/wk was 10,000 μCi (NBS 1953). Hence, 238 μCi/L [10,000 μCi/42L] 

would have produced 42.0 mrem/d; or 0.18 mrem/d would have been calculated from 1 μCi/L in urine. 

In the present model (ICRP 1997), 1.2 × 106 Bq/L results in 20 mSv/yr or 0.17 mrem/d dose is 

calculated from 1 μCi/L in urine. So the recorded doses for 1953 through 1972 are slightly higher per 

unit activity in urine than would be calculated using 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. 

Because the doses shown in Table 5-29 were based on only the non-zero recorded doses, they were 

judged to be most applicable to unmonitored workers at the 108B facility through 1954 and the 

Plutonium Recycle Test Reactor, 1960 through 1969. It is assumed that tritium doses at the reactors 

and fuel dissolution facilities would have been at most 10% of the doses in Table 5-29 [114]. The 

GSDs in Table 5-29 should also be used with a minimum GSD of 3. 

There was insufficient dose data in two periods, 1955 – 1960 and 1970 – 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 burn-up as a figure of merit for production (see Figure 1 in ORAUT 2007), average 

yearly production in 1955 through 1960 was 4 times greater than for 1949 through 1954. The average

of the median daily intakes for unmonitored reactor and fuel dissolution plant workers during 1949 

through 1954 was 3.2 × 105 pCi/d (one tenth of the average of the intakes in the fifth column in Table 

5-29). Four times this intake, 1.3 × 106 pCi/d, should be used for unmonitored reactor and fuel 

dissolution plant workers during 1955 through 1960. The largest GSD of 4.0 can be applied to the 

1955 through 1960 intakes. Production was essentially equal during 1970 and 1971 compared 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 was occurring during 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 1982-83. 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 [115] indicating 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 [116]. The 50-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, 1080, and 610 pCi/d, respectively [117]. N Reactor shut down in 

1987; PUREX ran intermittently in the late 1980s and was shut down in 1992 (See Section 2.2.3). 

The 1986 value can be applied to workers at those facilities through their respective shutdown dates

[118]. 

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 discussed below.

Drinking water for workers in the Hanford 400 Area, of which the principal facility was the Fast Flux 

Test Facility (FFTF), was obtained from the unconfined aquifer which 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 start-up of FFTF 

by a year or so. An analysis of the tritium concentrations in the wells providing 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 the DRs should use those data when available; for other 400 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 55 of 89

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 [119]. 

Operators, radiation monitors, and equipment maintenance workers at PUREX involved in the 233U 

campaigns might have had intakes without monitoring specific for 233U. The highest urinalysis result 

(4.39 pCi/d) from a batch of PUREX workers monitored in 1970 was used to estimate potential 

chronic intakes for the period 1965 through 1970 [120]. Because the material was absorption type F, 

urinary excretion equilibrates quickly so the same daily intake can be applied for 1 to 5 years chronic 

exposure. The daily intake producing 4.39 pCi/d uranium in urine was 16 pCi [121]. See the end of 

Section 5.2.5 for the contaminants that need to be added to this intake. These same workers might 

have been exposed to the thorium from which the 233U was extracted and which was recovered and 

recycled. See Section 5.3.4 for a description of monitoring for thorium. 

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. Note: ingestion intake of 65Zn 

should be assigned to all workers per OTIB-0039. 

100 Area Reactors 

• While operating, 1944–1987

•

: Use fission/activation products from OTIB-0039 and 131I and 

tritium intakes from Table 5-30. 

When not operating, 1988–present

300 Area Test Reactors, 1944–1978: Use fission/activation products from OTIB-0039 and tritium for 

PRTR from Table 5-30. 

: Use fission/activation products from OTIB-0039 only; no 

131I or tritium; no short-lived fission/activation products after 1988 for best estimates. 

400 Area, 1980–1994 (1978–1985 for tritium): Use fission/activation products (reactors) from OTIB0039 except use 24Na (F) as the additional radionuclide for best estimates. Note that this reactor 

continued to be used intermittently until 1994. Use tritium ingestion from drinking water per Table 5-

30. 

200 Area Fuel Separation Facilities

• T-Plant, 224-T, 1945–1956; B-Plant, 224-B, 1945–1952; REDOX, 224-S, 1952–1967; PUREX 

1956–1992

•

: Use fission/activation products, plutonium (M or S), and uranium (F) from OTIB0039, except can ignore 24Na from OTIB-0039 for best estimates. Use thorium intakes for 

PUREX 1965-70 from OTIB-0039 and 233U intakes for the same workers from Table 5-30 Use 

tritium and 131I intakes from Table 5-30. No short-lived fission/activation products after 1988 

for best estimates. 

B-Plant, 225-B, 1953–present

Plutonium Finishing Plants (various Z buildings and 233-S), 1945–present: Use plutonium 

intakes from OTIB-0039 (M or S). 

: Use fission/activation products from OTIB-0039, except can 

ignore 24Na and the “additional radionuclide” for best estimates. Use 65Zn ingestion-only 

intakes for 1953 – 1972. In other words exposure at these facilities was only to 137Cs and 90Sr, 

and 65Zn via the ingestion pathway. 

221-U, 224-U (UO3 Plant), 1952–1993: Use uranium type F intakes from OTIB-0039. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 56 of 89

108-B Tritium Extraction Facility, 1949–1956: Use tritium intakes from Table 5-30.

C-Plant (Hot Semiworks)

• 1949–1956

Table 5-30. Intakes for workers with external monitoring but essentially no bioassay [122]. 

: Use fission/activation products (fuel dissolution) from OTIB-0039/OTIB-0054, 

except can ignore 24Na for best estimates. Use plutonium (M or S) and uranium (F) from OTIB0039. 

Radioactive 

material Period 

Daily intake 

(pCi/d) Distribution

Pu Through 1988 See OTIB-0039 See OTIB-0039

U and Th Through 1988 See OTIB-0039 See OTIB-0039

Rn-220 in 3722 

or 3732 bldg.

1964 through 

1970

See Section 5.3.4 

for doses

Constant

Sr-90 Through 1988 See OTIB-0039 See OTIB-0039

Pm-147 See OTIB-0039 See OTIB-0039 See OTIB-0039

Zn-65 See OTIB-0039 See OTIB-0039 See OTIB-0039

Na-24 See OTIB-0039 See OTIB-0039 See OTIB-0039

Cs-137 Through 1988 See OTIB-0039 See OTIB-0039

Other 

fission/activation 

products

All See OTIB-0054 See OTIB-0054

I-131 (vapor) 1944 - 1959 2 E4 Lognormal, GSD =3

I-131 (vapor) 1960 - 1987 3.4 E3 Lognormal, GSD =3

1-131 (vapor) 1988 – present 0

Tritium (HTO) 1944 - 1948 6.6 E7 Lognormal, GSD = 3 108B

Tritium (HTO) 1944 - 1948 7 E6 Lognormal, GSD = 3 Reactors, fuel dissolution, 

rovers

Tritium (HTO) 1949 - 1954 From Table 5-29 Lognormal, from Table 

5-29

108B

Tritium (HTO) 1949 - 1954 0.1 times Table 5-

29

Lognormal, from Table 

5-29

Reactors, fuel dissolution, 

rovers

Tritium (HTO) 1955 - 1960 1.3 E6 Lognormal, GSD = 4 Reactors, fuel dissolution, 

rovers

Tritium (HTO) 1960 - 1969 From Table 5-29 From Table 5-29 PRTR

Tritium (HTO) 1961 - 1969 0.1 times Table 5-

29

From Table 5-29 Reactors, fuel dissolution

Tritium (HTO) 1970 - 1971 3.2 E5 Lognormal, GSD = 4 Reactors, fuel dissolution

Tritium (HTO) 1972 From Table 5-29 From Table 5-29 Reactors, PUREX plant 

Tritium (HTO) 1973-83 3.2 E5 Lognormal, GSD =4 N Reactor

Tritium (HTO) 1984 7.6 E3 Lognormal, GSD = 13 N Reactor, PUREX

Tritium (HTO) 1985 1.1 E3 Lognormal, GSD = 3.3 N Reactor, PUREX

Tritium (HTO) 1986-87 6.1 E2 Lognormal, GSD = 6.2 N Reactor, PUREX

Tritium (HTO) 1988-92 6.1 E2 Lognormal, GSD = 6.2 PUREX only

Tritium (HTO) 1978 - 1985 4.8 E 4 Lognormal, GSD = 3 Ingestion of drinking water 

in the 400 Area

U-233 (type F) + 

contaminants

1965, 1966, 

1970

16 Constant PUREX

D&D 

radiological 

workers

2002 – present See discussion, 

last paragraph in 

this section. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 57 of 89

• 1962–1967

300 Area Fuel Fabrication Facilities (303, 306, 313, 314, 333), 1944–1988: Use insoluble uranium 

intakes and thorium intakes from OTIB-0039. 

: Use fission/activation products from OTIB-0039/OTIB-0054 (fuel dissolution), 

except can ignore 24Na and the “additional radionuclide” for best estimates. Use 65Zn ingestion 

intakes only. 

209-E and 120 Critical Mass Laboratories, 1950–1986: Use plutonium intakes (M or S) from OTIB0039. 

308 Plutonium Fuels Pilot Plant, 1960–1990: Use fission/activation products (fuel dissolution), 

plutonium (M or S), and uranium (M or S), and 147Pm intakes from OTIB-0039/OTIB-0054. Use 131I 

intakes from Table 5-30. May ignore thorium and 24Na intakes for best estimates. 

324 Building, 1966–2001: Use fission/activation products (fuel dissolution) from OTIB-0039/OTIB0054. The fuel was well-aged before arriving so may ignore 24Na and other short-lived radionuclides 

in OTIB-0054 for best estimates, and may use 65Zn ingestion intakes only. 

325 Building, 1953–present: Use fission/activation products (fuel dissolution), plutonium (M or S), 

and 147Pm intakes from OTIB-0039/OTIB-0054. May ignore 24Na for best estimates; however, 

because of multiple uses of this facility, short-lived radionuclides may have been present past 1987. 

327 Building, 1953–1987: Use fission/activation products (fuel dissolution) and plutonium (M or S) 

intakes from OTIB-0039/OTIB-0054. May ignore 24Na for best estimates.

Tank Farms and Evaporators, 1944–present: Use fission/activation products (waste management) 

and plutonium (M or S) and americium intakes at 0.001 times the fission product intake (to account for 

a small alpha component in the waste) from OTIB-0039. May ignore 24Na for best estimates. May 

ignore 24Na, 65Zn, and the “additional radionuclide” intakes after 1988. 

D&D and Remediation, 2002 – present: For employees of the site’s remediation contractor whose 

job description indicates potential for exposure to airborne contamination, assign 40-DAC-hrs 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 claimants of the following 

(chose one): 260 pCi/d of either 90Sr (type F) 137Cs (type F), 152Eu (type M), 14C (particulate type M or 

S) or 0.26 pCi/d of 239Pu (Type M or S) [123]. 

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

separations facilities, except do not include thorium or 233U. 

Unknown facility: Use judgment based on any available information and OTIB-014. If can associate 

the work with the 100 Area, assume exposure at a reactor; if can associate with the 200 Area, then 

may have to assume the worst case, which would be the fuel separations facilities; if can associate 

the work 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. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 58 of 89

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.

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. 

[1] Bihl, Donald E. Battelle – Pacific Northwest National Laboratory (BPNNL). 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. BPNNL. 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. BPNNL. Principal Health Physicist. October 2003. 

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

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

[4] Bihl, Donald E. BPNNL. Principal Health Physicist. 

From a query on the REX database by the REX custodian. 

[5] Bihl, Donald E. BPNNL. Principal Health Physicist. 

The dates for HIE and ORE are from Wilson (1978). The other remarks are from personal 

experience and discussions with various program managers of the Radiological Records 

Program. 

[6] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

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

[7] Bihl, Donald E . BPNNL. Principal Health Physicist. October 2003. 

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

[8] Bihl, Donald E. BPNNL. 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. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 59 of 89

[9] Bihl, Donald E. BPNNL. 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. 

[10] Bihl, Donald E. BPNNL. 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. 

[11] Bihl, Donald E. BPNNL. 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. 

[12] Bihl, Donald E. BPNNL. 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.

[13] Bihl, Donald E. BPNNL. 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). 

[14] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

This is based on personal experience. 

[15] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

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

Laboratory Contract Technical Administrator. 

[16] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

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

contained in Carbaugh (1989). That two results were occasionally recorded for the same 

sample, because both splits were analyzed, is also evident from a query of the database. 

[17] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

The MDAs were extracted from various statements of work with 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. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 60 of 89

[18] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

The requirement to report total propagated uncertainty shows in the 1983 statement of work

(Battelle Northwest 1983). A query of the database shows that reporting of the uncertainty 

was sporadic in 1980 and became routine in 1981.

[19] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

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

including AEC-RL (1964), ERDA (1975), DOE-RL (1979), and Battelle Northwest (1983, 1985, 

1987, 1992). 

[20] Bihl, Donald E. BPNNL. 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 relocated. 

[21] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

This is based on discussions with PUREX engineers in the 1980s and from Sula et al. (1989). 

[22] Bihl, Donald E. BPNNL. 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 an internal memorandum

from D. Bihl to Program staff, “Unusual Plutonium Mixture” (Bihl 1992), which in turn was 

based on a description of the mixture in a more formal document, only a portion of which was 

found in the files. 

[23] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

In mixtures that are mostly americium by mass or where the americium atoms have been 

separated from the plutonium atoms, the americium behaves as americium, not as plutonium. 

[24] Bihl, Donald E. BPNNL. Principal Health Physicist. November 2004. 

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

[25] Bihl, Donald E. BPNNL. Principal Health Physicist. November 2004. 

No information was found to indicate whether the units were activity per sample or activity per 

liter during 1964. Guidance was required; therefore, the units used in 1967 to 1969 were 

assumed. 

[26] Bihl, Donald E. BPNNL. 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.

[27] Bihl, Donald E. BPNNL. Principal Health Physicist. November 2004. 

This is based on calculations by the author on the 1964 dataset in an attempt to determine the 

MDA of the procedure at that time. 

[28] Bihl, Donald E. BPNNL. 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 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 61 of 89

analysis. In addition, see Table 5-2, the discussion in the plutonium section, and 

Attribution 13. 

[29] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

This is based on personal experience, results of a query of the REX database, and research 

on some of the cases in the files. 

[30] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

This is based on personal experience. 

[31] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

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

laboratory, including AEC-RL (1964), ERDA (1975), DOE-RL (1979), and Battelle Northwest 

(1983, 1985, 1987, 1992). 

[32] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

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

laboratory, including AEC-RL (1964), ERDA (1975), DOE-RL (1979), Larson (1982), Battelle 

Northwest (1983, 1985, 1987, 1992), and Carbaugh (2003a, Appendix B). 

[33] Bihl, Donald E BPNNL. 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). 

[34] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

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

laboratory, including AEC-RL (1964), Lardy (1970), ERDA (1975), DOE-RL (1979), Larson 

(1982), and Battelle Northwest (1983, 1985, 1987, 1992). 

[35] Bihl, Donald E. BPNNL. 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. 

[36] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

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

laboratory, including AEC-RL (1964), Lardy (1970), ERDA (1975), DOE-RL (1979), Larson 

(1982), and Battelle Northwest (1983, 1985, 1987, 1992). 

[37] Bihl, Donald E. BPNNL. 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. 

[38] Bihl, Donald E. BPNNL. Principal Health Physicist. November 2006. 

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

dosimetry manual (Carbaugh 2003b) express the entity as ZrNb, 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 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 62 of 89

106RuRh. 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 ZrNb 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. 

[39] Bihl, Donald E. BPNNL. 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. Ms. Long 

concluded that uranium would be in a carbonate complex that most likely would be quite 

soluble (Long 1993). 

[40] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

This is based on personal experience. 

[41] Bihl, Donald E. BPNNL. 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 Mr. Bihl. 

[42] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

The rapid method was coded QUS for “quick uranium soluble.” Sula et al. (1989 Appendix C) 

refers to it as the “less sensitive method.” The kit code for the QUS was 7. 

[43] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

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

laboratory, including AEC-RL (1964), Lardy (1970), ERDA (1975), DOE-RL (1979), Larson 

(1982), and Battelle Northwest (1983, 1985, 1987, 1992). 

[44] Bihl, Donald E. BPNNL. 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 μg/L, 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 μCi/sample. Someone probably forgot to mark the special 

units on the uranium bioassay form. 

[45] Bihl, Donald E. BPNNL. 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. 

[46] Bihl, Donald E. BPNNL. 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. 

[47] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

This was explained to Mr. Bihl verbally during a regular contractual meeting between the 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 63 of 89

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. 

[48] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

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

laboratory, including AEC-RL (1964), Lardy (1970), ERDA (1975), DOE-RL (1979), Larson 

(1982), and Battelle Northwest (1983, 1985, 1987, 1992). 

[49] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

No use of strontium titanate at Hanford has been discovered.

[50] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

Personal files show an incident involving potential intakes of 147Pm in the 308 Building in 1971. 

Several workers were scheduled for routine 147Pm bioassay due to work with the material in 

the 222-S Building in 1964; other notes hint at earlier work with the material, probably smallscale chemistry development of processes used later in the 325 Building. 

[51] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

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

laboratory, including AEC-RL (1964), Lardy (1970), ERDA (1975), DOE-RL (1979), Larson 

(1982), and Battelle Northwest (1983, 1985, 1987, 1992). 

[52] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

This information is based on query of the database and review of the data. 

[53] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

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

laboratory, including AEC-RL (1964), Lardy (1970), ERDA (1975), DOE-RL (1979), Larson 

(1982), and Battelle Northwest (1983, 1985, 1987, 1992). 

[54] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

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

laboratory, including AEC-RL (1964), Lardy (1970), ERDA (1975), DOE-RL (1979), Larson 

(1982), and Battelle Northwest (1983, 1985, 1987, 1992). 

[55] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

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

laboratory, including AEC-RL (1964), Lardy (1970), ERDA (1975), DOE-RL (1979), Larson 

(1982), and Battelle Northwest (1983, 1985, 1987, 1992). 

[56] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

That 238Pu represents the majority of alpha activity in the 237Np source term has been tribal 

knowledge passed down through the Hanford Internal Dosimetry Program and is consistent 

with the assumption used at the Savannah River Site. However, proof in the form of analyses 

of the actual product has not been found. 

[57] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2005. 

This is based on queries of the REX database. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 64 of 89

[58] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2005. 

This is based on queries of the REX database. 

[59] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

This is based on discussion with the project manager, Dr. D. Fisher, a PNNL employee. 

[60] Bihl, Donald E. BPNNL. 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. BPNNL. Principal Health Physicist. October 2003. 

Mr. 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. BPNNL. Principal Health Physicist. October 2003. 

The earliest detection levels came from a slide in a training class on whole body counting at 

Hanford, circa 1964. The author was not stated, although it was most likely Earl Palmer, the In 

Vivo Monitoring Program Manager at the time (Author unknown 1964). Some of the detection 

levels were also shown, with calculations, in Earl Palmer’s 1960 Laboratory Record Notebook. 

Based on the calculations, Mr. 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. BPNNL. 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. BPNNL. 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. BPNNL. Principal Health Physicist. October 2003. 

The reporting of these radionuclides is evident from a query on the database. It was also 

obvious from hard-copy printouts of in vivo data from the first computer databases 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 Mr. 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. BPNNL. 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. See also 

ORAUT (2007a). 

[67] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

This is based on personal discussions with former in vivo counting program staff, H. E. Palmer 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 65 of 89

and G. A. Rieksts. Mr. Palmer was the in vivo counting program manager for many years, and 

Mr. 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. BPNNL. 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. BPNNL. Principal Health Physicist. October 2003. 

This is based on personal discussions with H. E. Palmer and G. A. Rieksts.

[70] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

This is based on personal experience and discussion with G. A. Rieksts. In addition, see for 

instance the 1992 “In Vivo Bioassay Statement of Work,” page 3.6 (Lynch et al. 1992). 

[71] Bihl, Donald E. BPNNL. 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. BPNNL. Principal Health Physicist. October 2003. 

This is based on personal experience and discussion with H. E. Palmer. 

[73] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2005. 

This is based on queries on the REX database for 131I results. 

[74] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2005. 

This is based on discussion with T. P. Lynch and G. A. Rieksts. 

[75] Bihl, Donald E. BPNNL. Principal Health Physicist. November 2006. 

This is based on calculations by Mr. Traub; however, similar values can be extracted from the 

graph in West (1965). 

[76] Bihl, Donald E. BPNNL. Principal Health Physicist. June 2005. 

The terms “detection limit’ or “less-than” during the 1960s appear to be most closely 

associated with the present concept of decision level. The MDA is approximately twice the 

decision level; therefore, the recommendation was made to assume that the MDA was twice 

the reported detection level. 

[77] Bihl, Donald E. BPNNL. Principal Health Physicist. November 2006. 

This is based on discussion with T. P. Lynch; however, this is an educated guess by Mr. 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 discussed in the text. 

[78] Traub, Richard. BPNNL. Principal Health Physicist. November 2006. 

This is based on calculations by Mr. Traub using MatlabTM. The calculations incorporated the 

latest ICRP biokinetic models with independent systemic biokinetics for the major progeny. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 66 of 89

[79] Bihl, Donald E. BPNNL. 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 were 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 generally improved over time. 

Therefore, 2 nCi is favorable to claimants. 

[80] Traub, Richard. BPNNL. Principal Health Physicist. December 2006. 

These were calculated by Mr. Traub using MatlabTM. The calculations incorporated the latest 

ICRP biokinetic models with independent systemic biokinetics for the major progeny. 

[81] Traub, Richard. BPNNL. Principal Health Physicist. December 2006. 

These were calculated by Mr. Traub using MatlabTM. The calculations incorporated the latest 

ICRP biokinetic models with independent systemic biokinetics for the major progeny. 

[82] Bihl, Donald E. BPNNL. Principal Health Physicist. December 2006. 

Because of the very high activities of fission products 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 same 

as normal fuel and thoron was exhausted out the canyon stack. 

[83] Bihl, Donald E. BPNNL. 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. BPNNL. 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 Palmer (1990). 

[85] Bihl, Donald E. BPNNL. 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. 

[86] Bihl, Donald E. BPNNL. 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. BPNNL. Principal Health Physicist. October 2003. 

This is evident from a review of various whole-body count results forms. 

[88] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

These changes were evident from review of the REX database. 

[89] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

Implementation of NEXEC is discussed in Lyon et al. (1996). Lynch (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. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 67 of 89

[90] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

This is based on personal experience and discussion with T. Lynch. 

[91] Bihl, Donald E. BPNNL. 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. BPNNL. Principal Health Physicist. May 2006. 

FFTF was sodium cooled so neutron activation of the sodium was a potential source of 

contamination. See Section 2.1.9. 

[93] Bihl, Donald E. BPNNL. Principal Health Physicist. March 2007. 

The principal 154Eu photopeak is at 1.27 MeV with 35% abundance; the photopeak 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] Bihl, Donald E. BPNNL. 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 with the 

bioassay laboratory (AEC-RL 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 (Battelle Northwest 1983) 

and in each statement of work thereafter. 

[95] Bihl, Donald E. BPNNL. 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. 

[96] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

Routine urine samples with volume less than 500 mL were rejected (code IS for insufficient 

sample) and the worker was asked to resample. This was a quality control measure designed 

to prevent samples that were grossly less than a 24-hour sample. The volume available in the 

collection containers (which varied over the years from 1.5L to 3L) and the limited amount of 

time for collection provided an upper limit to the uncertainty associated with 24-hour collection. 

Therefore, the factor of 2 was considered reasonable for the range of the uncertainty. Dose 

reconstructors use an overall uncertainty for internal dose estimates of a GSD of 3, so the 

uncertainty associated with sample collection is considered part of the overall uncertainty. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 68 of 89

[97] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

This is based on discussion with the in vivo monitoring program manager, T. Lynch. 

[98] Bihl, Donald E. BPNNL. Principal Health Physicist. October 2003. 

This is based on discussion with the in vivo monitoring program manager, T. Lynch. 

[99] Bihl, Donald. BPNNL. Principal Health Physicist. October 2003. 

This is based on discussion with the in vivo monitoring program manager, T. Lynch. 

[100] Bihl, Donald E. BPNNL. 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. Mr. 

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. BPNNL. Principal Health Physicist. October 2003. 

This is based on discussion with the in vivo monitoring program manager, T. Lynch, and G. A. 

Rieksts. 

[102] Bihl, Donald E. BPNNL. Principal Health Physicist. November 2004. 

This is based on input by R. Falk during review of Rev. 01 of this document. A GSD of 5 

covers a range of 2.3 orders of magnitude at the 95th-percentile confidence level. 

[103] Bihl, Donald E. BPNNL. 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. 

[104] Bihl, Donald E. BPNNL. 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 bioassay are not indicative of exposure. Therefore, after 

consultation with other ORAUT 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. 

[105] Bihl, Donald E. BPNNL. 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. 

[106] Bihl, Donald E. BPNNL. 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 in 1962 to 1969, so is was deemed to be favorable to claimants to apply 

the dose unmonitored workers in years before implementation of routine in vivo counts.
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 69 of 89

[107] Bihl, Donald E. BPNNL. 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. 

[108] Bihl, Donald E. BPNNL. 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 the 1987. 

[109] Bihl, Donald E. BPNNL. Principal Health Physicist. May 2006. 

This is based on query of the REX database. 

[110] Bihl, Donald E. BPNNL. Principal Health Physicist. May 2006. 

This is based on query of the REX database. 

[111] Bihl, Donald E. BPNNL. Principal Health Physicist. May 2006. 

This was calculated using IMBA. 

[112] Bihl, Donald E. BPNNL. 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. 

[113] Bihl, Donald E. BPNNL. Principal Health Physicist. Joseph Lochamy. Scientist. XCEL 

Engineering; and Richard Sparks. CDE Dosimetry Services. Principal Health Physicist. 

December 2006. 

The tritium dose data is from a query on REX by Mr. Bihl, a lognormal statistical analysis was 

performed by Mr. Lochamy in accordance with ORAUT-PROC-0095 (ORAUT 2006b), and a 

validation of statistical analysis was performed by Mr. Sparks.

[114] Bihl, Donald E. BPNNL. 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. 

[115] Bihl, Donald E. BPNNL. Principal Health Physicist. November 2006. 

This was calculated using IMBA. 

[116] Bihl, Donald E. BPNNL. Principal Health Physicist; Joseph Lochamy. Scientist. XCEL 

Engineering, Inc.; and Richard Sparks. CDE Dosimetry Services, Inc. Principal Health 

Physicist. December 2006. 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 70 of 89

The tritium dose data is from a query on REX by Mr. Bihl, a lognormal statistical analysis was 

performed by Mr. Lochamy in accordance with ORAUT-PROC-0095 (ORAUT 2006b), and a 

validation of statistical analysis was performed by Mr. Sparks.

[117] Bihl, Donald E. BPNNL. Principal Health Physicist. December 2006. 

These were calculated using IMBA. 

[118] Bihl, Donald E. BPNNL. 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.

[119] Bihl, Donald E. BPNNL. Principal Health Physicist. June 2005. 

These were calculated using IMBA. Ingestion of 40,000 pCi/d produces approximately 

1 mrem/yr to all organs. 

[120] Bihl, Donald E. BPNNL. 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 

also Section 5.2.5). 

[121] Bihl, Donald E. BPNNL. Principal Health Physicist. June 2006. 

This was calculated using IMBA.

[122] Bihl, Donald E. BPNNL. Principal Health Physicist. June 2006. 

This is a summary of information developed in the text. 

[123] Bihl, Donald E. BPNNL. 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 (Carbaugh 2003b). 

[124] Bihl, Donald E. BPNNL. 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. 02 Effective Date: 06/22/2007 Page 71 of 89

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GLOSSARY 

active

A term used in early writings at Hanford, circa 1940s and 1950s, to mean radioactive. 

Example, “production of acetylene from the active water, with subsequent measurement of the 

ionization caused by the tritium beta particle.”

aging

In the context of reactor fuel and mixtures of plutonium isotopes, the time since 241Am was 

separated from the plutonium mixture.

ashing 

A chemistry term meaning driving off all moisture in a sample; ashing can be 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. 

cooling

In the context of reactor fuel, the time since the fuel was removed from the reactor core. 

reliably detectable activity

Three standard deviations of the spectral continuum, plus has a peak discernable by the 

naked eye; used in in vivo counting circa 1980s. 

rep

Radiation Equivalent Physical. 83 ergs of energy from ionizing radiation per gram of tissue. 

This was redefined to be 93 ergs/g in the late 1940s or early 1950s. Later replaced by the rad.

simulated

In the context of urine sampling, collection of urine from about one-half hour before retiring to 

bed, through the sleep period, and for about one-half hour after rising for two consecutive 

nights to simulate a 24-hr sample or four consecutive nights to simulate a 48-hr sample. 

thoron

220Rn.
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 81 of 89

TABLE OF CONTENTS

SECTION TITLE

A.1 Codes Used in Bioassay and Internal Dose Records .............................................................

PAGE

82

A.2 Tolerance Dose and Tolerance Values at Hanford................................................................. 87

LIST OF TABLES

TABLE TITLE PAGE

A-1 Sample type codes................................................................................................................. 82

A-2 Bioassay reason codes .......................................................................................................... 82

A-3 Excreta sample kit codes ....................................................................................................... 83

A-4 In vivo body location codes .................................................................................................... 84

A-5 Excreta unit codes..................................................................................................................84

A-6 Excreta processing codes ...................................................................................................... 84

A-7 Excreta laboratory codes........................................................................................................ 85

A-8 Excreta no-sample codes....................................................................................................... 85

5-9 In vivo invalid result codes ..................................................................................................... 85

A-10 INTERTRAC mode-of-intake codes ....................................................................................... 86

A-11 INTERTRAC evaluation reason codes ................................................................................... 86

A-12 INTERTRAC source-of-intake codes...................................................................................... 86

A-13 INTERTRAC miscellaneous codes......................................................................................... 86

A-14 Special whole-body count resolution codes............................................................................ 87

ATTACHMENT A

CODES USED IN BIOASSAY AND TOLERANCE VALUES

Page 1 of 9 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 82 of 89

A.1 CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS 

(Adapted from Carbaugh 2003b) 

These codes apply to information contained 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 A-1. Sample type codes. 

Code Type of sample

B Blood

F Feces

S Sputum

T Tissue

U Urine

Table A-2. 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, prior to 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 following 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.

ATTACHMENT A

CODES USED IN BIOASSAY AND TOLERANCE VALUES

Page 2 of 9 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 83 of 89

Table A-3. 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-day 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 largevolume 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 prior to 

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-day period. 

Used for IPUL sampling. Designated sample date is 2 days 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. Prior to 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 prior to about 1990, but 

should have no impact on dose reconstruction).

ATTACHMENT A

CODES USED IN BIOASSAY AND TOLERANCE VALUES

Page 3 of 9 
Document No. ORAUT-TKBS-0006-5 Revision No. 02 Effective Date: 06/22/2007 Page 84 of 89

Table A-4. 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 A-5. 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/gram until 07-01-82 

μg/sample after 07-01-82

5 μCi/sample

6 μCi/L

7 nCi

8 μCi

Table A-6. 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 A-7. 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 A-8. Excreta no-sample codes. 

No-sample code Description

CN Kit not out. Sample kit not out at time of scheduled pickup.

CS Cancelled sample/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 A-9. 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 A-10. INTERTRAC 

mode-of-intake codes. 

Code Mode of intake

ABS Absorption

ING Ingestion

INH Inhalation

NON None (no intake)

UNK Unknown

WND Wound

Table A-11. 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 prior to Hanford employment

R Reevaluation

Table A-12. 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 prior to Hanford employment

Table A-13. 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 A-14. 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

A.2 TOLERANCE DOSE AND TOLERANCE VALUES AT HANFORD

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). Herbert 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/day. In 1936 the U.S. Advisory Committee on X-ray and Radium Protection 

reduced the tolerance dose to 0.1 R/day. 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/day for external X and gamma radiation

• 10–14 Ci/cc for radon in air of working rooms

• 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/week. In 1953, the 

NCRP developed maximum permissible amounts of certain radionuclides in the body and maximum 

permissible concentrations in air and water, published as National Bureau of Standards Handbook 52 

(NBS 1953). At that point tolerance values were replaced with Maximum Permissible Body Burdens 

and maximum permissible concentrations. 

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):

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External gamma and X radiation 0.1 R/day

External beta radiation 0.1 repa/day

Fast neutron radiation 0.02 rep/day

Slow neutron radiation 0.025 rep/day

Internal alpha radiation 0.01 rep/day

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. rep (roentgen equivalent physical) was a unit of dose from particulate 

radiation invented by Herbert Parker. It was a precursor to the rad. It was 

defined as an absorbed dose of 83 erg per gram of tissue and was later 

changed to 93 erg/g.

b. Assumed to refer to fission products [124]. 

c. Product was a euphemism for plutonium.

By March 1949, the Hanford limits had changed to incorporate the 0.3 rem per week guidance from 

the NCRP. Examples are quoted below from Patterson (1949), but these same air concentrations 

limits were discussed as Hanford limits in a lecture series by Parker (1947). The Patterson letter 

refers to these values as permissible tolerances.

External Radiation Limits

Whole body – 0.3 rem per week or whatever the National Committee on Radiation Protection 

recommends

Hands only – 1.0 rem per week or whatever the National Committee on Radiation Protection 

recommends

Internal Emitter Limits

0.3 rem per week to the significant organ or whatever the National Committee on Radiation Protection 

recommends

Drinking Water

Uranium – 100 μg/liter - believe this is high

Plutonium – 0.01 μg/liter – intend to revise to not more than 0.001 μg/liter

Mixed fission products – 0.1 μc*

Air Contamination Limits

/liter – to be changed when new figure provided by K. Z. Morgen’s 

subcommittee

Respiratory protection required

Uranium - >0.05 μg/liter

 

* Assumed to mean microcuries.

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Plutonium - >2 × 10-8 μg/liter to be revised perhaps to 2 × 10-9 μg/liter

Mixed fission products - >10-6 μc/liter

Other air contamination limits

Tritium – gas 0.1 μc/liter, vapor 0.01 μc/liter

Carbon – 0.02 μc/liter

Argon – 1.6 × 10-3 μc/liter

Iodine – 1.5 × 10-6 μc/liter

Xenon – 0.01 μc/liter

The tolerance values for air and drinking water were based on the dose to the significant organ and 

state-of-the-air 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 

(letter to file dated December 17, 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-hr exposure (Parker 1945). For the chronic exposure, Parker assumed a thyroid radiation 

tolerance of 1 R/day, 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 breathed per 8-hr work day 

was 8 × 106 cc. 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 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-hr exposure to be 1.2 × 10 11 Ci/cc. 

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