ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for the Hanford Site --

Occupational Internal Dose

Document Number:

ORAUT-TKBS-0006-5 

Effective Date: 11/24/2004

Revision No.: 01 

Controlled Copy No.: ________

Page 1 of 57

Subject Expert: Donald E. Bihl

Document Owner

Approval: Signature on File Date: 11/03/2004

Edward D. Scalsky, TBD Team Leader

Approval: Signature on File Date: 11/01/2004

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 11/01/2004

Richard E. Toohey, Project Director

Approval: Signature on File Date: 11/24/2004

James W. Neton, Associate Director for Science

Supersedes:

Revision No.: 00

TABLE OF CONTENTS

Section

Record of Issue/Revisions ...................................................................................................................

Page

4

Acronyms and Abbreviations ...............................................................................................................5

5.1 Occupational Internal Dose.........................................................................................................6

5.2 In Vitro Minimum Detectable Activities, Analytical Methods, and Reporting Protocols............... 13

5.2.1 Plutonium .........................................................................................................................13

5.2.2 Americium ........................................................................................................................18

5.2.3 Curium .............................................................................................................................21

5.2.4 Tritium..............................................................................................................................21

5.2.5 Uranium ...........................................................................................................................22

5.2.6 Fission Product Analysis .................................................................................................. 27

5.2.7 Strontium..........................................................................................................................28

5.2.8 Promethium......................................................................................................................29

5.2.9 Polonium ..........................................................................................................................30

5.2.10 Neptunium........................................................................................................................31

5.2.11 Other Limited-Exposure Radionuclides ............................................................................ 31

5.3 In Vivo Minimum Detectable Activities, Analytical Methods, and Reporting Protocols ............... 33

5.3.1 Whole Body Counters ...................................................................................................... 33

5.3.2 Chest Counters ................................................................................................................37

5.3.3 Thyroid Counters.............................................................................................................. 39

5.3.4 Head Counters and Other Counts .................................................................................... 40

5.3.5 General Notes about Items in the Database..................................................................... 40

5.4 Mixtures ....................................................................................................................................41
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5.4.1 Fission and Activation Products.............................................................................................. 41

5.4.1.1 Reactor workers............................................................................................................... 43

5.4.1.2 Separations Plants ........................................................................................................... 43

5.4.1.3 Waste Management Facilities (tank farms, evaporators, transfer lines)............................ 44

5.5 Interferences, Uncertainties ......................................................................................................45

5.5.1 Contamination of Samples ............................................................................................... 45

5.5.2 Uncertainties ....................................................................................................................45

5.6 Workers with No Confirmed Intakes..........................................................................................46

5.6.1 Special Consideration for Plutonium, Americium, and Thorium ........................................ 46

5.6.2 Worst Case Chronic Intakes............................................................................................. 47

5.7 Unmonitored Workers ...............................................................................................................48

References ........................................................................................................................................53

Glossary ............................................................................................................................................57
Effective Date: 11/24/2004 Revision No. 01 Document No. ORAUT-TKBS-0006-5 Page 3 of 57

LIST OF TABLES

Table

5.1-1 Air sample data ..................................................................................................................

Page

6

5.1-2 Codes and radionuclides associated with bioassay at Hanford........................................... 9

5.2.1-1 Routine plutonium urinalysis detection levels ................................................................... 14

5.2.1-2 MDAs for nonroutine Pu excreta analyses........................................................................ 15

5.2.1-3 Activity composition of Hanford reference weapons-grade plutonium 

mixture .............................................................................................................................16

5.2.1-4 Activity composition of Hanford reference fuel-grade plutonium 

mixture (12%)...................................................................................................................17

5.2.1-5 Activity composition of Hanford reference commercial power fuelgrade plutonium mixture................................................................................................... 18

5.2.2-1 Routine 241Am urinalysis detection levels.......................................................................... 20

5.2.2-2 MDAs for nonroutine 241Am excreta analyses................................................................... 20

5.2.3-1 Routine Cm urinalysis detection levels ............................................................................. 21

5.2.3-2 MDAs for nonroutine Cm excreta analyses....................................................................... 21

5.2.4-1 Routine tritium urinalysis detection levels ......................................................................... 22

5.2.5-1 Radiological characteristics of Hanford uranium mixtures................................................. 23

5.2.5-2 Impurities in recycled uranium at Hanford ........................................................................ 24

5.2.5-3 Inhalation class for Hanford uranium compounds............................................................. 24

5.2.5-4 Routine uranium urinalysis detection levels ...................................................................... 26

5.2.5-5 MDAs for nonroutine uranium excreta analyses ............................................................... 26

5.2.6-1 Routine fission product urinalysis detection levels ............................................................ 28

5.2.7-1 Routine 90Sr urinalysis detection levels............................................................................. 29

5.2.8-1 Routine 147Pm urinalysis detection levels.......................................................................... 30

5.2.9-1 Routine 210Po urinalysis detection levels........................................................................... 31

5.2.9-2 MDAs for nonroutine 210Po excreta analyses.................................................................... 31

5.3.1-1 Routine whole body counting detection levels .................................................................. 34

5.3.1-2 Mean body burdens of 137Cs from fallout in the United States .......................................... 37

5.3.2-1 Routine chest counting detection levels............................................................................ 38

5.7-1 Intakes for workers with external monitoring but no bioassay ........................................... 49

5.7-2 Beta/gamma radionuclide chooser for unmonitored workers ............................................ 51
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 10/02/2003 00-A New document to establish TBD for occupational 

internal dose – section 5. Initiated by Edward D. 

Scalsky.

10/15/2003 10/15/2003 00 First approved issue. Initiated by Edward D. 

Scalsky.

Draft 07/13/2004 01-A Major revision 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 non-routine 

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. 

Initiated by Edward D. Scalsky

Draft 09/17/2004 01-B Incorporates internal review and NIOSH 

comments. Initiated by Edward D. Scalsky.

Draft 10/29/2004 01-C Incorporates additional internal review and NIOSH 

comments. Initiated by Edward D. Scalsky.

11/24/2004 11/24/2004 01 Approved issue of Revision 01. Initiated by 

Edward D. Scalsky.
Effective Date: 11/24/2004 Revision No. 01 Document No. ORAUT-TKBS-0006-5 Page 5 of 57

ACRONYMS AND ABBREVIATIONS

CEDE Committed Effective Dose Equivalent

CF Commercial Fuel

DU Depleted Uranium

GeLi Lithium drifted Germanium (detector)

GI Gastro Intestinal

GOK God Only Knows

HIE Hanford Internal Exposure (database)

HPGe High Purity Germanium (detector)

ICRP International Commission on Radiological Protection

INEEL Idaho National Environmental Engineering Laboratory

KPA Kinetic Phosphorescence Analysis

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

MPBB Maximum Permissible Body Burden

MPC Maximum Permissible Concentration

NCRP National Council on Radiation Protection and Measurements

NU Natural Uranium

ORE Occupational Radiological Exposure (database)

PNL Pacific Northwest Laboratory

PNNL Pacific Northwest National Laboratory

RDA Reliably Detectable Activity

REX Radiological Exposure (database)

RU Recycled Uranium

TPU Total Propagated Uncertainty

TRU Transuranic

TTA Thenoyl trifluoroacetone

UST United States Testing Company
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5.1 OCCUPATIONAL INTERNAL DOSE

Technical Basis Documents and Site Profile Documents are general working documents that provide 

guidance concerning 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 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 facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act of 2000 (42 U.S.C. § 7384l (5) and (12)).

When the first reactor was started 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 of personnel monitoring was with the 

Medical Department. For this time frame, air sampling data, environmental data and incident data will 

likely be the only information available to use for recreating personnel exposure. Information about air 

sample results was highlighted in monthly reports of the Health Instrument Department. These were 

very brief summaries, mostly highlighting problems; often no values were listed, indicating air 

concentrations were below concern. Air sample data from the reactors were almost never listed in 

these reports; the radiation protection emphasis at the reactors seemed to be external dose and 

effluents in the water. The records show that high air concentrations at the other facilities prompted 

use of respiratory protection. Table 5.1-1 summarizes the air sample data found so far. 

Table 5.1-1. Air sample data. 

Year Facility Max concentration (μCi/cm3)

Most sampler 

concentrations (μCi/cm3)

1943 Building 305 Test Reactor

1944 B Reactor

D Reactor

T Plant

1945 T,U,B Canyon Bldgs 8E-12 Pu <8E-13 Pu

T,U,B Concentrator Bldgs (224T,U,B) 2E-11 Pu <2E-12 Pu

D Reactor

B Reactor

F Reactor

231-Z One at 8E-11 Pu, very temporary and 

area immediately placed on mask; most 

highs were about 8E-12 Pu or less.

<8E-13 Pu

300 Area Labs 6E-12 Pu <8E-13 Pu

Metal Fab. Bldgs. 1E-9 Unat <2E-10 Unat

1946a T,U,B Canyon Bldgs. 4E-12 Pub

T,U,B Concentrator Bldgs. 

D Reactor

B Reactor

F Reactor

231-Z 4E-11 Pu 8E-13 Pu

300 Area Metal Fab. Bldgs. 2E-9 Unat <1E-10 Unat

3706 Bldg. 6E-12 Pu, 5E-10 Unat <2E-12 Pu

200 W Laundry 1E-11 Pu 4E-12 Pu

a. Based on monthly reports for July and September through December only.

b. Excluding one incident in B canyon involving only two workers, for which special urine samples were obtained. 
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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 prior to when the NCRP introduced the terms 

Maximum Permissible Body Burden and Maximum Permissible Concentration. Air concentration 

tolerance levels were based on dose rates to significantly-impacted organs. Examples at Hanford 

were 0.01 rep per day from alpha-emitters, 1 rep per 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 for total accumulation of plutonium in the body of 0.5 μg. 

The radiation protection program was directed toward preventing workers from acquiring a tolerance 

dose or being exposed to tolerance air concentrations. Respiratory protection was supposed to be 

worn if an air concentration exceeded the tolerance level or if a tolerance level was anticipated to be 

exceeded for a given job. The tolerance air concentration for plutonium in 1945 was 4 x 10-11 μCi/cm3

(Cantril 1945). Additional discussion on tolerance levels is provided in Attachment D. 

Air sample data were not routinely reported in the Health Instrumentation Section monthly reports for 

the reactors. There were three exceptions in 1946: 1) a high air sample of 3x 10-9 μCi/cm3 beta 

activity when a gasket around a thimble blew, 2) 6.5x10-9 μCi/cm3 beta activity at 100 D – listed as the 

highest value for the year, 3) 6x10-8 μCi/cm3 beta activity for a task at 100 F but workers were wearing 

respirators. 

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

(canyon) buildings during this time. Concern was for 131I uptake for workers who entered the canyons, 

such as crane operators. No information was reported as to instruments used, MDAs, or results. The 

1946 annual summary report stated that nothing significant was detected in any of the thyroid scans, 

without stating what was considered significant. However, the tolerance level for 131I in air had been 

established in October 1945 as 1x10-7 μCi/cm3 (Cantril 1945) based on a permissible equilibrium 

amount in the thyroid of 2 μCi. Based on other statements in the monthly reports, it is reasonable to 

assume that scans showing thyroid burdens over 2 μCi would have been considered significant. 

Air concentrations for which respiratory protection was required were established at least by 1947 and 

probably earlier. The values were 1 x 10-12 μCi/cm3 (alpha) for plutonium and 1 x 10-9 μCi/cm3 for 

fission products (assumed to mean particulate beta-emitters) (Parker 1947, Patterson 1949). 

The Health Instrument Section Monthly reports do mention contamination spreads in the reactor 

buildings, 231-Z, concentrator buildings, and uranium metal fabrication shops during these years, so 

intakes undoubtedly were occurring. Large intakes of plutonium would have been detectable in later 

years when bioassay was available; but some level of chronic intake during this period is a 

reasonable assumption. Chronic intakes of uranium in the metal fabrication shops should also be 

assumed. The tolerance air concentration for uranium machining was 1.5x10-4 μg/cm3

, which 

converts to 1.1x10-10 μCi/cm3

. Most air sample data for the “Metal Fabrication Buildings” were simply 

listed as less than that level. 

Because there is little information concerning intakes in the years prior to implementation of routine 

bioassay programs, the following default assumptions should be made unless there is better 

information in the worker’s file. All of the values below should be considered the modes of triangular 

distributions with zero minimums and maximums of twice the modes. Although it is unlikely that all 

workers were actually exposed to air concentrations at just under the respiratory protection required 

levels for 40 hours per week (basis for the mode), it is also possible that workers were exposed to 

greater concentrations for short periods of time. The maximum values account for 

• short term exposure to higher air concentrations
Effective Date: 11/24/2004 Revision No. 01 Document No. ORAUT-TKBS-0006-5 Page 8 of 57

• the possibility that the respiratory protection was not always effective

• uncertainty in the date the respiratory protection requirement was implemented. 

Reactor workers, 1944-48: Assume inhalation intakes of fission products based on the respiratoryprotection-required concentration, i.e., 1 x 10-9 μCi/cm3

, for 40 hours per week. Assuming a breathing 

rate of 1.2 m3

/hr, this produces a chronic inhalation intake of 9,600 pCi per work day or 6,600 pCi per 

calendar day. In addition, assume an ingestion intake of 200 pCi per calendar day, based on 

guidance in “Estimation of Ingestion Intakes” (NIOSH 2004). [(0.2)(1 x 103 pCi/m3

)] Although reactor 

workers were normally exposed to just activation products, anecdotal information indicates that the 

integrity of the fuel cladding was not always good and exposure to fission products during refueling or 

handling of spent fuel was possible. Use Table 5.7-2 to choose which fission or activation product 

radionuclide to apply. 

Separations plants, 1944-46: Assume inhalation intakes of plutonium alpha (6% Pu mix, see Table 

5.2.1-3) absorption type M at 1x10-12 μCi/cm3 for 40 hours per week. As described above, this 

produces a chronic intake of 9.6 pCi per work day or 6.6 pCi per calendar day. In addition, assume 

an ingestion intake of 0.2 pCi per calendar day. Assume intakes of fission products based on the 

respiratory-required concentration, i.e., 1 x 10-9 μCi/cm3

, for 40 hours per week. This produces a 

chronic inhalation intake of 9,600 pCi per work day or 6,600 pCi per calendar day plus an ingestion 

intake of 200 pCi per calendar day. Use Table 5.7-2 to choose which radionuclide. These chronic 

intakes would apply from either the first day of work for the worker or the start up of the plant 

(December 1944 for T Plant and April 1945 for B Plant). Also assume chronic inhalation intake of 131I 

of 7.5 x 10+5 pCi per day SR-1 vapor, which will produce equilibrium thyroid burdens at just under the 

2-μCi tolerance level (ICRP 1997, Table A.6.20). The iodine intake, excluding environmental 

radioiodine, would only apply to workers that entered the canyons or perhaps the main stack sampling 

buildings, but it’s unlikely the dose reconstructor (DR) will be able to differentiate these workers from 

general workers at the separations plants, in which case the DR should apply the intake to all workers 

at these plants with job descriptions that imply some risk of intake. Ingestion of 131I was unlikely. 

231-Z, 1945-46: Assume intakes of plutonium alpha (6% Pu mix, see Table 5.2.1-3) absorption type 

M or S at 1x10-12 μCi/cm3 for 40 hours per week. This produces an inhalation intake of 9.6 pCi per 

work day chronic intake or 6.6 pCi per calendar day plus an ingestion intake of 0.2 pCi per calendar 

day. 

300 Area uranium fabrication buildings (313, 314), 1944-47: Assume chronic inhalation intakes of 

1900 pCi per day natural uranium (1300 pCi per calendar day) and ingestion intakes of 60 pCi per 

calendar day (based on assumed average air concentrations of 2x10-10 μCi/cm3 and other parameters 

described above). (See Section 5.2.5 for the isotopic composition.) Assume absorption type S or 

Type M depending on the organ of concern. 

Laundry 1944-46: Assume chronic inhalation intakes of plutonium alpha (6% Pu mix, absorption 

Type M) of 38 pCi per work day (27 pCi per calendar day) and ingestion intakes of 0.8 pCi per 

calendar day based on an air concentration of 4 x 10-12

μCi/cm3, 8 hr per day, and other parameters 

described above. This intake assumes exposure to the soiled laundry for the entire day. If interviews 

indicate that part of the time was spent at the washing station or handling the cleaned laundry, the 

intake may be reduced by the ratio of hours spent handling clean laundry/8.

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 
Effective Date: 11/24/2004 Revision No. 01 Document No. ORAUT-TKBS-0006-5 Page 9 of 57

methods, routine urine sampling and analysis for plutonium was initiated in 1946. Urinalysis for 

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

products started in this time frame 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. Additionally, 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. But the records as a whole list a wide spectrum of radionuclides that 

were monitored and an even longer list of codes used to identify either 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 whose 

tasks “might” have exposed them to lots of different sources, for instance, radiation monitoring 

technicians. 

Table 5.1-2 provides a fairly exhaustive list of codes for analyses that can be encountered 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 just show the IPA code with a reason for not obtaining a result. 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 xray/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. The 

code GOK (God only knows) shows on in vivo count hardcopy records during the 1960s and 70s. 

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

spectrum from NaI-based whole body counters. 

Table 5.1-2. Codes and radionuclides associated with bioassay at Hanford. 

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
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Code Description Comment

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

Table 5.1-2 (Continued). Codes and radionuclides associated with bioassay at Hanford.

Code Description Comment

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 God only knows 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 Am241 Excreta scheduling code

IPIU Isotopic Pu, isotopic U Excreta scheduling code 

IPS Isotopic Pu and Sr Excreta scheduling code

IPSA Isotopic Pu, Sr tot & Am241 Excreta scheduling code; Sr tot means radiostrontium by gross 

beta

IPSR Seq Pu isotopic Sr-total Excreta scheduling code for isotopes of Pu and radiostronium

IPU Isotopic plutonium Excreta scheduling code

IPUB Plutonium isotopic, Pu241 Excreta scheduling code; Pu-241 separate anal. by beta counting

IPUBA Plutonium isotopic, Pu241, 

Am241

Excreta scheduling code
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Code Description Comment

IPUL Low level isotopic Pu Pu-238 and Pu-239 using a 10,000 minute count

IRA Radium isotopic Excreta scheduling code

IR192 Iridium 192 Excreta scheduling code

ISCP Sequential Sr 90 Ce Pm Excreta scheduling code

ISPEC Gamma spectroscopy Excreta scheduling code

ISR Sr isotopic Excreta scheduling code

ITH Thorium isotopic Excreta scheduling code

ITPAC Seq isotopic Pu, Cm & 

Am241

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

Table 5.1-2 (Continued). Codes and radionuclides associated with bioassay at Hanford. 

Code Description Comment

NP239 Neptunium 239

PB210 Lead 210

PB212 Lead 212

PM147 Promethium 147

PO210 Polonium 210

PR144 Praseodymium 144

PU Plutonium alpha Total alpha from Pu isotopes after separation

PUMIX Plutonium alpha Total alpha from Pu isotopes and Am-241

PU238 Plutonium 238

PU239 Plutonium 239 When pertaining to excreta samples, it’s actually Pu-239+240

PU240 Plutonium 240

PU241 Plutonium 241

PU242 Plutonium 242

QUS U Quick Uranium Soluble; excreta scheduling code for elemental U

QUS 1 U Same as QUS

QUS 2 U Same as QUS

RA224 Radium 224

RA225 Radium 225

RA226 Radium 226

RA228 Radium 228

RH106 Rhodium 106

RND Radon daughters

RU103 Ruthenium 103

RU106 Ruthenium 106

S 35 Sulfur 35

SB124 Antimony 124

SB125 Antimony 125

SCP Sequential Sr-total Ce Pm Excreta scheduling code
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Code Description Comment

SM153 Samarium 153

SR Strontium Total radiostrontium by beta counting

SR 89 Strontium 89

SR 90 Strontium 90 When pertaining to excreta samples, Sr-90 by yttrium ingrowth

TAC Total actinides

TC 99 Technetium 99

TH227 Thorium 227

TH228 Thorium 228

TH230 Thorium 230

TH232 Thorium 232

TH234 Thorium 234

TL208 Thallium 208

U Elemental uranium

URAN Elemental uranium

U DEP Depleted uranium

U NAT Natural uranium

U 233 Uranium 233 See uranium discussion in text

U 234 Uranium 234 Actually U-234 + 233, but usually U-234 

U 235 Uranium 235

U 236 Uranium 236

U 238 Uranium 238

UMIX Uranium mix Total uranium, used for intakes not bioassay

UMS U 235 U 236 U 238 U 234

Table 5.1-2 (Continued). Codes and radionuclides associated with bioassay at Hanford.

Code Description Comment

US U

XX 0 Isotope will have no result

ZN 65 Zinc 65

ZR 95 Zirconium 95

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 (as in at Hanford or other site),

• nature of intake, and

• mode of intake.

Tables listing and explaining these codes are provided in Attachment D.
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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 principal the REX database has all the information 

from the previous databases, but as stated above there may be isolated situations where some data 

never got into a database or some data did not get transferred from one database to another. 

There is another anomaly found in the results circa 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. 

Home sampling began very early in the program (1946) and has continued throughout the history of 

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

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

Electric) occurred in 1965. At that time the DOE-Richland Office established a contract for in vitro 

analyses with the United States Testing Company, which built and operated a commercial low-level 

radiochemistry lab 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. 

Except for a period between 1990 to 1992, despite a series of competitive procurements, in vitro 

analyses have been performed in the same facility since 1965. However, due to buyouts and 

mergers, the name of the laboratory has changed in the following sequence: United States Testing, 

International Technology Analytical Services, Quanterra Environmental Services, and Severn Trent 

Laboratories (present).

Battelle defaulted the contract with United States Testing in June 1990, and subsequently routine 

samples were collected and frozen (Lyon 1991, Lyon 1992). Between September and November 

1990 temporary contracts/agreements were established and samples were being analyzed at the 

following laboratories: Los Alamos National Lab (plutonium), TMA-Norcal (strontium), PNL-Analytical 

Chemistry Lab [325 Building] (tritium), and Westinghouse Hanford Company [222-S Building] 

(elemental uranium). In February 1991, IT Analytical Services commenced analyses for plutonium, 

americium, curium, and isotopic uranium. Los Alamos National Lab was replaced by Oak Ridge 

National Lab and Reynolds Electric and Engineering Company at the Nevada Test Site (plutonium) in 

April 1991. The contract with IT Analytical Services replaced the former contract with United States 

Testing, but the other labs continued to process samples until the backlog was worked off. So the 

work at the temporary labs was finishing up during late 1991 through early 1992 with the last results 

being received in March 1992. 

5.2.1

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

plutonium started in September 1946. The first plutonium bioassay analysis consisted of lanthanum 

fluoride precipitation and thenoyl trifluoroacetone (TTA) extraction and gross alpha counting. 

Plutonium
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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 

5.2.1-1. The definition of “detection level” no doubt changed over the years, but the levels in Table 

5.2.1-1 fit reasonably with the concept of limit of detection or MDA. For example, the Wilson history 

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

confidence levels. Occasionally recovery, counting, etc., allowed detection levels to be as low as 

0.028 dpm and for a short period, a level of 0.027 dpm was reached and used as the detection level. 

. . . This practice [of recording lower detection levels] was discontinued and the more conservative 

0.05 dpm was used routinely even though lower levels were possible part of the time.” 

Prior to October 1983 the recorded value was the total alpha activity from plutonium so would have 

included activity from 238Pu, 239Pu, and 240Pu. Any 241Pu or 241Am present in the urine would not have 

been accounted for by the recorded results. 

The results may 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 x 10-8 was recorded for results for which plutonium was not detected (one 

half of the nominal 0.05 dpm MDA). This method of recording was used through 1974. In 1975 the 

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

plutonium was not detected. 

Table 5.2.1-1. Routine plutonium urinalysis detection levels. 

Period

MDA, 

dpm/sample

Decision level, 

dpm/sample Measured quantity

Prior to June 1949 0.96a 0.66 Total Pu alpha

6/1949 to 11/1952 0.33 “

12/1952 to 1/27/53 0.18 “

1/28/53 to 3/26/53 0.15 “

3/27/53 to 11/06/53 0.05 “

11/07/53 to 12/04/53 0.07 “

12/53 to 4/55 0.057 “

5/55 to 8/55 0.027b “

9/55 to 9/55 0.04b “

10/55 to 9/30/83 0.05c 0.025c “

10/01/83 to 12/31/83 0.035 Each Pu-238, Pu-239

1/02/84 to 4/88 0.02 “

5/88 to 5/90 0.02 0.01 “

6/90 to 11/91 0.03 0.015 “

11/91 to 4/2000 0.02 0.01 “

5/2000 to 8/2001 0.02 Xb + 2.05x TPUd “

9/2001 to present 0.02 2 x TPUd “

a. Estimated from data in Table 1 in “Bioassay at Hanford” (Healy 1948)

b. The values are probably closer to a decision level than an MDA. Recommend using 0.05 

dpm as the MDA. 

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

0.025. But if net activity above background and above 0.025 was detected the actual amount 

was recorded.

d. Xb is mean of blanks and TPU is total propagated uncertainty.
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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 minute counting time has been 

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

was rare. 

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 

INEEL and found to work very well. 

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. 

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

Generally the lab performed slightly better than the contractual MDA, but the true MDA varied slightly 

over time and the contractual MDA was a reliable estimate. 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. 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. Initially a fixed value of 0.01 dpm/sample was used for all results, being one half the 

nominal MDA. The decision level was allowed to become sample-specific based on the total 

propagated uncertainty in 2000, and an adjustment was made to the formula in 2001. 

The MDAs listed in Table 5.2.1-1 apply to routine and priority processing of urine samples. Fecal 

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

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

MDAs for these samples are provided in Table 5.2.1-2.

Table 5.2.1-2. MDAs for nonroutine Pu excreta analyses. 

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. 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 labs but the MDA was not 

established, probably about 0.2-0.5 dpm/sample considering state-of-the-art of those labs.

Normally, fecal sampling was 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 
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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, 1993; Lyon et al 1988; Lyon et al 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. 

Except for a few standards in radiochemistry laboratories, 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 with the purpose of 

producing 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 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, the target mixture was the weapons grade mixture. Tables 5.2.1-3 and 5.2.1-4 lists the 

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

240Pu mixtures (from Carbaugh 2003). 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 concerning the composition of the source of the intake. For instance, only in rare, large 

intakes was 241Pu measured as part of the intake so the activity of that isotope is almost never 

available. 241Am at 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. So unless 241Am was measured or there is other information about the intake, there 

may 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) the ratios in Table 5.2.1-3 should be 

used. However, lacking any helpful information about the intake, an assumption of 10-year-old 

fuel grade plutonium mixture would be claimant-favorable and reasonable. For intakes since 

about 1996, 20-year-old fuel grade mixture could be assumed. 

Table 5.2.1-3. 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) 238Pu 8.56E-03 8.23E-03 7.91E-03 7.60E-03 7.31E-03 7.03E-03 6.75E-03 239Pu 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02 240Pu 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02 241Pu 8.24E-01 6.48E-01 5.09E-01 4.00E-01 3.15E-01 2.48E-01 1.95E-01 242Pu 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06 241Am 0 5.83E-03 1.04E-02 1.39E-02 1.66E-02 1.87E-02 2.03E-02 239+240Pu 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
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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) 239+240Pu:241Am NA 12.2 6.87 5.13 4.28 3.80 3.50 239+240Pu:238Pu 8.33 8.67 9.01 9.38 9.74 10.1 10.5 241Pu:239+240Pu 11.6 9.09 7.15 5.62 4.42 3.48 2.73

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

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

Pu alpha:241Am NA 13.6 7.62 5.68 4.73 4.19 3.84 241Pu: Pu alpha 10.3 8.15 6.43 5.07 4.01 3.17 2.50

a. Time since separation of 241Am from the Pu mix.

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

reactors (Nuclear Waste Vitrification Project). The 324 and 325 Buildings in the 300 Area were 

involved. Commercial fuel rods have a much higher degree of “burnup,” and the ones at Hanford 

were characterized by much more 241Pu and nominally 26% 240Pu. Table 5.2.1-5 provides the activity 

characteristics of the commercial fuel used in the Nuclear Waste Vitrification Project. In addition, the 

Plutonium Finishing Plant sometimes recycled plutonium from other DOE sites. This material would 

be rich in 241Am. Plutonium from the West Valley commercial reprocessing site is also stored at 

Hanford. But unless the records concerning the specific intakes being investigated have evidence of 

these unusual mixtures, the default mixtures mentioned above should be used.

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

sets are not compatible. For a first approximation, for curve fitting in IMBA and POC determination in 

IREP, the Pu alpha data can be treated as 239Pu and for the post 1983 data, the 239Pu and 238Pu 

values can be summed and treated as 239Pu. However, the intakes of 241Pu and 241Am must be 

included in the dose determination for input into IREP. If the POC is marginally close to the 50% 

criterion, then the total Pu alpha intake (as 239Pu) should be split out into actual intakes of 238Pu and 

239Pu because the dose conversion factors are not the same. 

Most plutonium at Hanford was in moderately soluble form, e.g. nitrates, which can be modeled as 

inhalation (absorption) type M. But many forms were possible over the years, especially metal and 

oxides. Even material, such as old contamination, that was originally in soluble form has a tendency 

Table 5.2.1-4. 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) 238Pu 1.71E-02 1.64E-02 1.58E-02 1.52E-02 1.46E-02 1.40E-02 1.35E-02 239Pu 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.25E-02 240Pu 2.72E-02 2.72E-02 2.72E-02 2.72E-02 2.72E-02 2.71E-02 2.71E-02 241Pu 3.09E+00 2.43E+00 1.91E+00 1.50E+00 1.18E+00 9.29E-01 7.30E-01 242Pu 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06 241Am 0 2.19E-02 3.89E-02 5.22E-02 6.24E-02 7.03E-02 7.63E-02 239+240Pu 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 239+240Pu:241Am NA 3.64 2.05 1.53 1.28 1.13 1.04 239+240Pu:238Pu 4.67 4.86 5.05 5.24 5.46 5.69 5.90 241Pu:239+240Pu 3.87E+01 3.05E+01 2.40E+01 1.88E+01 1.48E+01 1.17E+01 9.16

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

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

Pu alpha:241Am NA 4.39 2.46 1.82 1.51 1.33 1.22 241Pu: Pu alpha 31.9 25.3 20.0 15.8 12.5 9.91 7.83

a. Time since separation of the 241Am from the Pu mix.
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Table 5.2.1-5. Activity composition of Hanford reference commercial power fuel-grade plutonium

mixture. 

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) 238Pu 1.71E-01 1.64E-01 1.58E-01 1.52E-01 1.46E-01 1.40E-01 1.35E-01 239Pu 3.41E-02 3.41E-02 3.41E-02 3.41E-02 3.41E-02 3.41E-02 3.41E-02 240Pu 5.90E-02 5.89E-02 5.89E-02 5.89E-02 5.89E-02 5.88E-02 5.88E-02 241Pu 1.34E+01 1.05E+01 8.28E+00 6.51E+00 5.12E+00 4.03E+00 3.17E+00 242Pu 1.97E-04 1.97E-04 1.97E-04 1.97E-04 1.97E-04 1.97E-04 1.97E-04 241Am 0 9.49E-02 1.69E-01 2.26E-01 2.79E-01 3.04E-01 3.31E-01 239+240Pu 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 239+240Pu:241Am NA 0.981 0.551 0.411 0.344 0.305 0.281 239+240Pu:238Pu 0.544 0.568 0.589 0.612 0.636 0.664 0.688 241Pu:239+240Pu 144 113 89.1 70.0 55.1 43.3 34.1

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

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

Pu alpha:241Am NA 2.72 1.49 1.08 0.857 0.770 0.689 241Pu: 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 Pu mix.

to oxidize when left in contact with air, such as old contamination. Oxides, metal, and old 

contamination should be treated as inhalation type S. 

If nothing is known about the chemical form of the plutonium, then either type M or S can be used. If 

there are sufficient bioassay data to determine the type by curve fitting, use the best fit; otherwise use 

the type that maximizes the dose to the organ of concern. 241Am that is a component of plutonium 

contamination should be modeled in the lung the same as the plutonium matrix in which it has 

ingrown. In other words the americium should be treated as 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, because 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), waste tanks, transfer lines, and a whole operation in 

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). The ICRP recommended absorption type for americium is M. 

Americium

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

americium excreta analysis in the 1948 report by Jack Healy, “Bioassay at Hanford;” no mention in a 

1954 memo, “Bioassay Annual Report,” that lists numbers of urinalyses for plutonium, fission 

products, and uranium; no mention in a compilation of bioassay procedures, given a title of “Bioassay 

Procedures and Analysis (Old Bioassay Bible),” but no author or editor, dated April 10, 1961. 

Hanford bioassay records show 41 urinalyses for 241Am in 1964 collected from 19 workers. No 

americium urinalyses were recorded immediately before 1964 and none in 1965 or 66. The samples 

were collected over a short time period. The first samples, twelve of them, were collected on August 

25 or 26. The last sample was collected on October 29. The results range from 1.3 x 10-6 to 
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6.4 x10-6. No units are given, although the magnitude of the results suggest the units are μCi/sample 

or μCi/L, which is consistent with the reporting style later in the 1960s. It is not clear whether the units 

are per sample or per liter. Per liter was used in 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 the 

other; hence, assuming the units are μCi/L is reasonable.

One set of four samples for one worker were 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. Hence, 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. 

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 x 10-6 μCi/L and a 95th percentile of 2.3 x 10-6 μCi/L. Based on 

this analysis the MDA for 241Am analyses in 1964 was assumed to be 4.6 x 10 -6 μCi/L.

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 6 workers 

that were involved in the same potential intake accident in 222-S on May 2, 1967. There are internal 

dosimetry evaluations in the workers’ files documenting the accident and bioassay results. Many of 

the urine sample results were listed as <5.41 x 10-7 μCi/L so it is assumed that this was the MDA for 

the analysis. This is the same number that was handwritten on the side of an unpublished article, 

“Evaluation of Internal Depositions of Americium Using Bioassay Samples,” by RC Henle dated May 

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

A memo to file from John J. Jech, Senior Development Engineer in the Personnel Dosimetry Services, 

dated September 1969, states that per a telephone conversation with Matt Lardy at the U. S. Testing 

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

described, in the personal interview, the procedure as DDCP extraction to a planchet and gross alpha 

counting. A letter was found from Matt Lardy to Harold Larson, manager of Personnel Dosimetry 

Services, dated March 27, 1974, stating that the new limit for 241Am in urine is 0.1 pCi/sample at the 

90% confidence limit. This limit was still listed in a statement of work with U.S. Testing in 1979 and 

again in 1982, although it was therein stated as 0.2 dpm/sample. In the laboratory statement of work 

for a new contract starting October 1983, the detection level was listed as 0.04 dpm per 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 that the results are 241Am is claimant favorable. 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 HASL-300 manual and 

commonly referred to as the “RICH-RC-50-80” method. 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. The MDA 

in the 1985 statement of work was listed as 0.02 dpm/sample consistent with the change to alpha 

spectrometry, and it has stayed there to the present. Presently, Eichrom TRU column exchange is 

used for the separation of the americium for urine; however, the MDA is the same. 
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Table 5.2.2-1 summarizes what has been uncovered concerning 241Am MDAs for routine urinalysis. 

Table 5.2.2-1. Routine 241Am urinalysis detection levels. 

Period

MDA, 

dpm/sample

Decision level, 

dpm/sample

1964 4.6E-6 μCi/L Anything detected

1967 to 6/1969 5.4E-7μCi/L

7/1969 to 2/1974 2.0 Anything detected

3/1974 to 10/1983 0.2 “

10/1983 to 9/1985 0.04 “

10/1985 to 05/1988 0.02 “

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 x TPUa

9/2001 to present 0.02 2 x TPU

a. Xb is the mean of the blanks and TPU is total propagated uncertainty.

The MDAs listed in Table 5.2.2-1 apply to routine and priority processing of urine samples. Fecal 

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

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

MDAs for these samples are provided in Table 5.2.2-2. These analyses may have been used 

because of suspected intakes of pure 241Am (such as the famous 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 U.S. Testing. On rare occasions for a serious intake, samples were analyzed for 241Am using a 

low-energy photon detector, prior to any chemistry. This technique came into existence in 1986 or 87. 

Its detection level was about 5 dpm/sample. Generally, the LEPD result was just used as a rapid 

indicator, and a more accurate result was obtained by wet chemistry/alpha spectrometry days later.

Table 5.2.2-2. MDAs for nonroutine 241Am excreta analyses. 

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 SOW) 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.

e. 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 labs but the MDA was not established, probably about 0.2-0.5 

dpm/sample considering state-of-the-art of those labs.
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5.2.3

The curium isotopes of concern were 242 and 244, although sources of curium at Hanford were 

minor, usually calibration sources or as minor constituents in an actinide mixture. The curium and 

americium procedure were 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. The MDAs were not always identical with 241Am, however. Routine urinalysis MDAs for 

curium are provided in Table 5.2.3-1 and non-routine excreta analyses are provided in Table 5.2.3-2.

Curium 

Table 5.2.3-1. Routine Cm urinalysis detection levels. 

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 x TPUa

9/2001 to present 0.02 2 x TPU

a. xb is the mean of the blanks and TPU is total propagated uncertainty.

Table 5.2.3-2. MDAs for nonroutine Cm excreta analyses. 

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)cd 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.

5.2.4

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

mentioned at all in the Wilson history of personnel dosimetry (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 (and apparently an instrumentation expert as well). That procedure was 

based on “production of acetylene from the active water, with subsequent measurement of the 

ionization caused by the tritium beta particle” (Healy 1949). No detection level was mentioned in that 

letter, but one was mentioned in an internal memo from Herbert M. Parker to A.B. Greninger, dated 

January 1950, that referred to the acetylene method for urinalysis and provides a sensitivity of about 

1.2 μCi/L in water (Parker 1950). However, that method apparently didn’t work well because Healy 

stated in a 1951 letter to H.F. Schultz at Los Alamos 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 

Tritium
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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 

previously mentioned interview with Matt Lardy, Mr. Lardy stated that liquid scintillation counting of a 

1-ml aliquot of raw urine has been used since U.S. Testing was awarded the bioassay contract in 

1965. 

Tritium intakes were accounted for as part of external dose until about 1986 or 87, when they were 

entered in the dose database as an internal dose. 

Basically tritium was not a major source of radionuclide exposure for large numbers of workers at 

Hanford. 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 by-product 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 exposure was assumed to 

be chronic during the exposure period, unless a very large acute intake was known to occur. 

Tritium was referred to as P-10 in the 1950s. The main source of tritium in the 1950s was 108-B, also 

called the P-10 Plant, which started in August 1949.

Very little data on MDAs has been discovered. A 1964 letter to the PRTR Radiation Monitoring 

personnel (McConnor, 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-thans. The 1965 statement of work with US Testing shows an MDA of 1 μCi/L (which is 

consistent with the MDA at Savannah River Site throughout the 1950s). Table 5.2.4-1 provides MDAs 

for routine tritium urinalysis as best has been compiled to date. From 1978 to present the MDAs were 

obtained from statements of work with the bioassay laboratory; the MDAs and time periods prior to 

that are guesses. 

Table 5.2.4-1. Routine tritium urinalysis 

detection levels. 

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 

1961was 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 that was also called recycled uranium (RU). Small numbers of researchers may 

have experimented with more enriched uranium at different times, e.g. metallurgy on commercial 

grade fuel, but such exposure would have been to small groups for limited periods. Table 5.2.5-1 

Uranium 
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provides the default uranium mixtures (Carbaugh 2003). Generally, 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) . Recycle uranium also had impurities build up and track with the 

uranium over time. Impurities can be approached in two ways, representative levels based on 

averages of several measurements at different times and upper limits based on tolerance 

specifications (e.g., not to exceed). Both of these approaches are given in Table 5.2.5-2. 

Table 5.2.5-1. Radiological characteristics of Hanford uranium mixtures. 

Uranium mixture

Weight percentage a, b Natural (NU) Depleted (DU) Recycled (RU) Commercial fuel (CF)

234U 0.0057 0.0005 0.0082 0.0300

235U 0.7204 0.2500 0.9700 2.9600

236U Negligible Negligible 0.0680 Negligible

238U 99.2739 99.7500 98.9500 97.0100

Specific constituent activity in mixture (uCi/g, nCi/mg, or pCi/ugc

234U 0.3563 0.0313 0.5125 1.8750

235U 0.0156 0.0054 0.0210 0.0639

236U Negligible Negligible 0.0440 Negligible

238U 0.3336 0.3352 0.3325 0.3260

Total 0.7054 0.3718 0.9099 2.2649

Specific constituent activity in mixture (dpm/ug)c

234U 0.7909 0.0694 1.1378 4.1625

235U 0.0345 0.0120 0.0465 0.1419

236U Negligible Negligible 0.0977 Negligible

238U 0.7405 0.7441 0.7381 0.7236

Total 1.5659 0.8254 2.0200 5.0281

Constituent fraction of total uranium activity in mixture

234U 0.5051 0.0840 0.5632 0.8279

235U 0.0221 0.0145 0.0230 0.0282

236U Negligible Negligible 0.0484 Negligible

238U 0.4729 0.9014 0.3654 0.1439

Total 1.0000 1.0000 1.0000 1.0000

a. NU, DU, and 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.

The rigorous radiation protection barriers and procedures designed to prevent intakes of plutonium 

and fission products were not, in general, applied to work with uranium. Hence, exposure to uranium 

in the major uranium facilities was considered chronic exposure until 1992.

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 
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Table 5.2.5-2. Impurities in recycled uranium at Hanford. 

Constituent Maximum alloweda Observed rangeb Reference levelc

Plutonium 10 ppb U <1 - 2 ppb U 0.4 nCi Pu-alpha/gU

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

Thorium 750 ppm U 8 - 10 ppm U 5 pCi 232Th/g U

99Tc Not established 3 - 4 ppm U 0.2 uCi 99Tc/g U

103,106Ru <20 uCi/lb U <6 uCi/lb U 40 nCi 106Ru/g U

95ZrNb <10 uCi/lb U <4 uCi/lb U 20 nCi 95ZrNb/g U

Other gamma emitters <2 uCi/lb U 0.09 - 0.75 uCi/lb U Negligible 

a. From UO3 Plant operating specifications, OSD-U-185-0001 (Thompson 1986).

b. From analysis of uranium lots 88-1, 88-2, 88-3 that were processed in 1988, and lots 93-01, 93-02, 93-

03, 93-04, and 93-05, processed in 1993.

c. A reference level is chosen for determining bioassay monitoring needs and for use as an initial 

assumption in evaluating intakes. The use of the reference levels is expected to result in a slight 

overestimate of dose compared to levels actually observed in 1988.

the major uranium handling facilities. Results are shown in Table 5.2.5-3. Because the relationship 

between the old lung fluid studies and the ICRP 66 absorption types is not established, Table 5.2.5-3 

also shows recommended absorption types for intakes from the listed facilities, which should be used 

unless person-specific data are available. These absorption type assumptions should be applied to 

the impurities as well. 239Pu can be assumed for the plutonium alpha impurity.

Table 5.2.5-3. 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 or U carbonate (assumed form after discharge to the soil) Md

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

b. Because the conversions from the solubility studies to the ICRP absorption types are not exact, the dose reconstructor may use the 

same percentages for D to F, W to M, etc. or may 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. Letter Report to Monte J. Sula from Darrell R. Fisher, January 20, 1986.

d. Cooke and Holt 1974. 

A note about sampling of UO3 Plant workers: Because chemical toxicity was the principal concern for 

uranium exposures at 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 

circa early 1980s. Change over should be clear in the records. 

The Friday/Monday sampling scheme was also used in 1962-63 for 313 and 314 Building workers. 

The Wilson history 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 various refinements over the years including some upfront chemistry on the raw urine, 

until about 1991, when it was replaced by kinetic phosphorescence analysis (KPA) (Lardy 2003). 
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[Note: Mr. Lardy said about 1990 but other evidence indicates late 1991.] A 1970 letter 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; another with extraction (after wetashing) 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. A third method was also listed; this was 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.” (Lardy 1970) The 

detection limit was given as 0.5 dpm/sample. A 1989 description of the chemistry was wet-ashing 

with HCl and extraction with hexone. 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. Generally, 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. Generally, personnel working in the production facilities were monitored 

by the elemental analysis, whereas Pacific Northwest Laboratory workers were monitored by the 

isotopic analysis because of the wide scope of research projects that occurred over the years. 

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 were reported as 234U from 1994 to present 

unless it was specifically determined that the worker was exposed to 233U. Work with 233U did occur at 

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

bioassay. So unless specifically mentioned in an intake investigation report, assume 233U results 

since 1983 are actually 234U.

233U was handled at 231Z Building in the mid 1960s as a special project, maybe extending into the 

early 1970s. This project involved thorium campaigns at PUREX, separation of the 233U, and 

shipment to 231Z Building. No details about this work have been uncovered as yet, such as isotopic 

purity. Because of the time frame, bioassay must have been for elemental uranium, at least until 

about 1970. If so, because of the high specific activity of 233U, the bioassay MDA would have been 

only about 90,000 dpm/L. Hopefully, something better was done, and the worker’s record might show 

that, but as yet the specific bioassay used for the 233U project has not been discovered. 

Table 5.2.5-4 summarizes the routine urinalysis detection levels and Table 5.2.5-5 summarizes 

nonroutine detection levels. 

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. 

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 to 238U ratio can be used to distinguish depleted uranium from natural uranium, 

but, considering uncertainties in analytical results, that ratio can not be used to distinguish recycled 
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Table 5.2.5-4. Routine uranium urinalysis detection levels. 

Period

Elemental Isotopic

MDA,

μg/L

Decision level, 

μg/L 

MDA, 

dpm/sample

Decision level, 

dpm/sample

Prior through 1948 Not specifically mentioned NA NA

1948 through 1949 10 Anything detected NA NA

1950 through 1969(a) 4 (b) NA NA

1970 through 1974 0.5 0.5

1974 through 1981 0.4

1982 through 9/1983 0.05 – 0.25 (0.1 

most probable)(c)

10/1983 through 

12/1983

0.03 0.5 (d) 0.035

1/1984 through 8/1985 0.03 0.02

9/1985 through 6/1990 0.03/0.5(e) 0.2(f) 0.02

6/1990 through 

10/1991

0.2/0.5(e) 0.2(f) 0.03 0.15/0.015(g)

11/1991 through 

present

0.06/0.5(e) 0.2(f) 0.02 0.15/0.010(g)

a. Estimated time period based on 1954 and 1970 letters.

b. Values were reported well below the 4 μg/L value so either the MDA was thought to be lower than that value or a 

decision level of 2 μg/L was being applied.

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

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.

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.

Table 5.2.5-5. MDAs for nonroutine uranium excreta analyses. 

Period

Analysis 

type

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

Emergencya,d Expedite Priority Emergencya,d Expedite

1980b Elemental 10 μg/L

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 10/1991 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 labs but the MDA was not established, probably about 0.2-0.5 

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

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 50 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, 

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

Hence, 0.2 μg/d was established in 1985 and continues to be used at present as the environmental 
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decision level for exposures to natural or recycled uranium. Only urinary excretions 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 that 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’s unlikely there will be data available to prove it. 

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. 

5.2.6

Fission product urinalysis was the method used to monitor for intakes of fission products until whole 

body counting was implemented 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. So data prior to 

1948 should be considered unreliable and should be ignored (see guidance in 5.1 instead). The 

procedure initiated in 1948 was to add Sr carrier to the aluminum oxide solution for the plutonium 

procedure, then precipitate La hydroxide. This procedure was shown to extract the rare earths and 

strontium with yields ranging from 90% for Ce to 23% for Sr. The dried planchet was counted for beta 

activity with an approximate detection level of 30 dpm. (Healy 1948, Wilson 1987) The same 

procedure was in use in 1954 with the addition of a Ce 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 gasflow, beta proportional counter in November 1958 which resulted in increased counting efficiency. 

The new detection limit was stated as 1.4 x 10-5 μCi/sample, based on the counting efficiency of 90Sr. 

“Gross fission products” are also mentioned in the 1970 letter from Matt Lardy at US Testing 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 5.2.6-1 summarizes the detection levels for the fission product urinalysis as best has been 

uncovered. 

Fission Product Analysis

It’s 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 that a worker might theoretically have been 

exposed to. The procedure separated and counted radionuclides of alkaline earths and rare earths, 

such as strontium, yttrium, barium, lanthanum, cerium, europium, and promethium. It did not account 

for radionuclides of ruthenium, cesium, zinc, cobalt, manganese, niobium, or zirconium. 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. The abundances of all the fission products, 
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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 

Table 5.2.6-1. Routine fission product 

urinalysis detection levels. 

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 E-5 μCi/ sample.

c. Listed in the bioassay contract but probably 

not used; replaced by whole body counting 

and 90Sr urinalyses.

contamination was inhaled or ingested. See 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. So it was apparently being used as a 90Sr bioassay. The records show fission product 

analysis being used this way until early 1964. The same workers show actual 90Sr analysis results 

starting in 1965, probably starting with the new contract with US Testing. 

5.2.7

Records of 90Sr urinalyses, both routines and specials, begin showing up 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 memo, dated July 1963, documenting 

discussions between the Analytical Laboratories and Internal Dosimetry clarifying logistics of handling 

these samples and reporting 90Sr results. There are handwritten notes on this memo indicating that 

the detection level is about 20 dpm. Nevertheless, the database records show fission product 

urinalyses being used into1964 and 90Sr urinalyses apparently starting in 1965. The value of 1.67 x 

10-5 μCi/L (37 dpm/L) is frequently entered in the database during 1965 and 1966 and seems to be 

the reporting level. This is consistent with a draft of the first contract with UST (the official one 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. The 1970 Lardy to Corley letter 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 US 

Testing shows an “analytical limit” (defined as ±25%) at 50 dpm/sample and a reporting level of 2 

dpm/sample. These values show again in the 1978 statement of work except the analytical limit is 

defined as ±100%. A 1979 letter from Bob Robinson (PNL Internal Dosimetry) to R.B. Swoboda (US 

Testing bioassay supervisor) requests changes for 90Sr urinalyses so that the analytical limit (±100%) 

be lowered from 50 dpm/sample to 5 dpm/sample, the reporting level be increased from 2 to 5 

dpm/sample, and that an emergency analysis capability be added 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. But in the new contract starting October 1983 the 

Strontium
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detection limit was listed as 2.0 dpm/sample, and it stayed at that value until 1992 when it was raised 

to 10 dpm/sample. However, the procedure stayed the same throughout this period and the true MDA 

probably held at about 2 dpm/sample.

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. 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 5.2.7-1 summarizes the routine 

urinalysis detection levels for 90Sr procedure.

Table 5.2.7-1. Routine 90Sr urinalysis detection levels. 

Period MDA or MDC

Prior to 1965 May have been available but MDA not known

1965 to 1969 37 dpm/L

1970 to 1974 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. 

All strontium results at Hanford should be considered absorption type F. It is claimant-favorable to 

assume that 90Sr and total radiostronium results are 90Sr even though 89Sr may be present. Because 

the 90Sr urinalyses method coincided in time with whole body counts, which would signal intakes of 

other fission products, 90Sr urinalysis results should represent only strontium intakes (i.e., not be used 

as an indicator for other fission products unless they were detected in whole body counts). The 

exception would be 147Pm, which probably tracked with the strontium through the various processes. 

See 5.4 for discussion of mixtures. 

5.2.8

Hanford was involved in the manufacture of heat sources using 147Pm. The time 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 and as late as 1971 in the 308 Fuels Laboratory. Also animal studies 

were conducted with 147Pm as part of research to develop a human biokinetic model for the behavior 

of promethium in the body. A small human volunteers study using 143Pm was conducted in 1967 or 

1968 (Palmer et al 1969). 

Promethium

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

separation chemistry then calcining. There was also one mention of cold-pressed, sintered Pm2O3 for 

heart implants. According to ICRP 68, the nitrate form should be considered absorption type M and 

the oxide form absorption type S.
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In the 1960s, 147Pm sample results were reported as, for urine - μCi/L, for feces – μCi/kg, which is 

different than most radionuclides, which were reported as per sample. From 1974 forward, the results 

appear to be reported as per sample. Table 5.2.8-1 lists the 147Pm minimum detection levels at 

various times.

Table 5.2.8-1. Routine 147Pm urinalysis detection levels. 

Period MDA or MDC

Prior to 1965 May have been available but MDA not known

1965 to 1969 37 dpm/L (1.67E-5 μCi/L)

1970 to 1973 22 dpm/L (1.0E-5 μCi/L)

1974 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

b

11/1991 to present 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 1.67x 10-5 μCi/kg. 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 in 1983 – 

1990s.

Only one description of the procedure was found, and that same procedure showed up in documents 

dated 1970, 1974 and 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 <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 

liquid scintillation spectrometer. 

5.2.9

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

210Po was found in the files circa 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 in which it was stated that the 210Po starts in the process in the 

soluble form but is converted to the insoluble form. However, U.S. Testing 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. There also was work with 210Po in the 325 Building that 

started in June 1972 and was slated “to run for 2-3 years.” 

Polonium

The procedure developed for 210Po by U.S. Testing 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 degrees C for 2 hours. The silver film was counted by alpha 

proportional counting. Fecal samples were first wet-ashed in concentrated nitric acid and peroxide 
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then treated the same as urine samples. Sometime between 1968 and 1974, the silver foil was 

replaced by copper foil and alpha spectrometry counting had replaced proportional counting. 

Detection limits for routine urinalysis are shown in Table 5.2.9-1 and for nonroutine excreta bioassay 

in Table 5.2.9-2. 

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

Table 5.2.9-1. Routine 210Po urinalysis detection levels. 

Period MDA

3/1968 to 1973 5.4E-7 μCi/L

1974 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. Not likely 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 5.2.9-2. MDAs for nonroutine 210Po excreta analyses. 

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. 

living around Hanford had been conducted. ICRP 23 (1975) indicates that excretion levels differ 

between smokers and nonsmokers, and provides the following estimated excretion values: urine, 

smokers: 0.065 pCi/d, nonsmokers: 0.011 pCi/d; feces, smokers: 3.3 pCi/d, nonsmokers: 3.2 pCi/d. 

These values were based on only 7 subjects, however, and even so, the fecal excretion ranged from 

1.7 to 6.4 pCi/d. 

If there are person-specific baseline values for urine or fecal excretion of 210Po, those should be used 

to subtract from later results. If not, then the ICRP 23 values above should be used; if smoking status 

is not known, use the values for nonsmokers. 

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. 

Neptunium 

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 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 

Other Limited-Exposure Radionuclides
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throughout the history of Hanford. The following discussion, addressing 14C, 232Th, radon, 90Y, 227Th, 

227Ac, and 32P, is not likely comprehensive.

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

samples were brought from the operating reactors to the 3731 Building for destructive testing. No 

information has been uncovered yet as to what bioassay if any was done. 14C 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. Baseline samples were obtained from each worker 

because natural excretion levels had not been established. ICRP 68 and 71 assign 14C in organic 

compounds to class SR2, which has not been modeled in IMBA yet. If a claimant appears to have 

been exposed to organic 14C, special consultation with the IMBA programmers may be necessary to 

determine an acceptable method to determine the dose. 

PUREX ran thorium campaigns in the 1960s and early 1970s. In terms of grams or curies, the thorium 

campaigns were small compared to the normal separation of plutonium. 232Th was irradiated to 

produce 233Th, which decays to 233U. Although called a thorium campaign, it was the 233U that was 

separated at PUREX and transported to 231-Z for experiments. Hence, the most likely source of 

intake was the 233U during the loading out and transportation. Thorium exposure was more likely at 

the 3732 Building where the powdered thorium fuel targets were fabricated, which apparently 

contaminated the building with thorium “fines.” 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 

DAC-hrs. The urinalysis MDA was stated to be 0.1 dpm/sample. 

There was a radon generator used for animal studies in the 108F Building and 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. 

Some unusual radionuclides were isolated in the 325 Building for nuclear medicine studies in the mid 

to late 1990s. One of these projects isolated 90Y from 90Sr and packaged and shipped the 90Y to 

various 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 that 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.

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 ceased after only a few milkings. Only a couple of 

researchers were involved. 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.” 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 uncovered if such a case is encountered. 
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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).

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 plate on all six sides, and graded shielding on the inner 

surfaces (lead, cadmium, copper) (Wilson 1987, Roesch et al 1960). The counting geometry was a 

chair configured to simulate a one-meter arc. The original count time was 20 minutes which was 

reduced to 10 minutes in October 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, so apparently the 

difference between the systems was not great enough to warrant republishing the MDAs. So the 

MDAs shown in Table 5.3.1-1 are the only MDAs found for the 1960s and 70s, and apparently were 

meant to apply generally to the various whole body counters in operation during this period. 

Whole Body Counters

Shortly after the chair counter in the “Iron Room” became operational, an entirely new design called 

the shadow shield counter was developed. The shadow shield consisted of a bed shielded on the 

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

that was also shielded by lead except for 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 located nearer the worksites on the Hanford site. The mobile trailer also had a thyroid 

detector and a wound counter. The mobile, 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 listing 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 1961-63 results added 46Sc, 51Cr, and 59Fe to the list. 

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. A listing of MDAs was found 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 was put in operation in 1985 and 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 person being counted. There are four 9.375-in-diameter-by-4-in.-thick detectors and one 
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11-in.-diameter-by-4-in.-thick detector, the latter being located behind the thoracic region. Count time 

is 200 seconds. 

In July 1989, a coaxial HPGe scanning array was developed and is still in operation today. For this 

system the person lies on a bed in a shielded room and the detector array moves under the bed. The 

configuration of this system, in terms of number and size of the detectors, has changed many times. 

It started as four 68% HPGe detectors; one of the detectors was replaced with a 120% detector in late 

1995; in May 1997 the system was upgraded to include seven detectors including three 120% 

detectors. When a 4-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 minutes; however, 20-minute count times are used as confirmation of an initial count with 

Table 5.3.1-1. Routine whole body counting detection levels.

a

Period Nuclide MDA (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 0.5

Cr-51 15 15

Mn-54 2.0 10

Fe-59 4.0 10

Co-60 2.0 10

Zn-65 3.0 3

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 2

Ce-144 100 100

1985-86

Na-22 1.5 1.5

Mn-54 3 3

Fe-59 6 6

Co-60 3 5

Zr-95 3 3

Ru-106 12 12

Eu-154 4.5 4.5

Cs-137 3 6

1987

Co-60 3 3

Cs-137 3 3

No changes for other radionuclides. Anything detected is reported.

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

decision level.

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

1995-10/1999e

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
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Period Nuclide MDA (nCi) Reporting level (nCi)

Na-24 1 Every result

Pr-144(Ce-144) 230 Every result

Other radionuclides Anything detected

10/1999 to presente,f

Co-60 1.25 Every result

Cs-137 1.3 Every result

Eu-154 3.75 Every result

Other radionuclides Anything detected

a. Nominal MDAs based on whatever phantom was available at the time period, 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.

c. See also discussion on thyroid detectors.

d. Based on 99% confidence of detection.

e. Least sensitive of many options throughout the period. Much better sensitivities were available using the HPGe system.

f. Physical configurations stayed essentially the same but ABACOS software introduced changes to methodology for determining MDAs 

and decision levels. 

detectable activity. Consequently, the database will usually show a 10-minute count and a 20-minute 

count on the same day or a few days later if the first count had detectable activity (excluding 40K or 

medical radionuclides). 

The Bi-214 radon progeny and the Tl-208 thoron progeny 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. 

Even with the background contribution subtracted, activity levels of Bi-214 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” (DIL) were implemented in October 1995 

for Bi-214 and Tl-208 measurements using the coaxial germanium counting system. The DIL values 

were 6.47 nCi for Bi-214 and 0.75 nCi for Tl-208 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, Bi-214 and Tl-208 

results have not been reported unless specifically requested.

Results above the DIL value were reported to the Internal Dosimetry Program staff for an investigation 

into the potential for exposure to radium or thorium. Such exposure was rare at Hanford. The 

investigation usually simply involved contact with the worker’s supervisor to determine if exposure to 

radium or thorium was possible. If an occupational exposure was suspected, the worker’s file will 

have a letter addressing the potential exposure. 

The first mobile counter stopped being used at various onsite locations 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. The sensitivity of the detector was 

comparable to the standup counter at the Whole Body Counter Facility. The use of this facility was 

infrequent and it was discontinued in August 1995. 
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From 1960 to 1983, four radionuclides were reported routinely: 24Na, 40K, 65Zn, 137Cs. 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. Both the reactors onsite and many of the cities surrounding the site used the Columbia River 

for sanitary water, although the 24Na intakes came mostly from the drinking water at the reactors 

because its short physical half-life greatly reduced the activity in city water supplies (Brady 1964). 

Hence, intakes of 24Na and 65Zn should be considered chronic ingestions, except for the very highest 

body burdens that were statistically different from the general background in Hanford workers. 

Guidance will need to be developed to distinguish acute inhalations from chronic ingestions. Chronic 

ingestion intakes would have been occurring prior to implementation of whole body counting. likely 

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). Net counts in a fifth region of 

the spectrum was also 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 there were any counts left over in the low energy 

region. If there were sufficient counts left over, then they would have 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. (GOK stands for God Only 

Knows.) 

If the hardcopy form (In-Vivo Counter Results) shows the “traces of xxx invalidate routine calculation” 

statement, then some radionuclide other than the standard four was detected; often this was 60Co. 

The activity of that radionuclide may or may not 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 5.3.5 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 uncovered in the records, and may not have been developed so 

long as the 137Cs measurement didn’t exceed 1% of a MPBB. The following guidance may be used 

however. 

• The 137Cs intake should be considered occupational if the same whole body count detected 

other fission or activation products (excluding 65Zn or 24Na). It should also be considered 

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 was done and the record clearly shows that the intake was due to a 

nonoccupational source, then the 137Cs may be disregarded. 

• NCRP Report No. 94 (NCRP 1987) provides mean body burdens of 137Cs for the United States 

for the years most likely to produce interference with occupational whole body count results. 

Those values are listed in Table 5.3.1-2. 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 given in Table 

5.3.1-2, the 137Cs result may be assumed to be due to fallout.
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Table 5.3.1-2. 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 Report No. 94.

In early to mid 1990s, 214Bi and 208Tl were reported if they were detected above background levels. 

The background levels were established using counts from an unexposed sample of workers for the 

“standup” counter and from long empty-room counts for the coaxial germanium counter. These background levels were subtracted from the worker’s counts in the respective regions of interest. 

However, results were sometimes above the decision level (especially for 214Bi in winter during

temperature inversions); if the results were still above the environmental decision level after a shower 

and a second count, they were reported to the database. There were no known sources of 226Ra at 

Hanford at the time and minimal exposure to thorium, if any, but the reporting of these natural 

radionuclides allowed for a small investigation to be conducted to ensure that the worker was not 

exposed to radium or thorium (usually just a phone call to the supervisor). The results of the 

investigation should be in the worker’s dosimetry file, probably on the In Vivo Exam Questionnaire. 

5.3.2

In 1967 the original large NaI detector in the Iron Room started to also be used for chest counting, 

with emphasis on uranium workers. The detector was placed directly over and nearly in contact with 

the chest region with the worker in the supine position. Count time was 30 minutes. MDAs were 

determined to be 6.7 nCi for “U natural,” presumably based on 234Th, 0.15 nCi for 235U, and 0.33 nCi 

for 241Am. However, in the next year a new counting room was built, called the Lead Room, 

specifically for chest counting. It was outfitted with four 5-in.-diameter by 0.375-in-thick NaI detectors, 

located two in front and two in back of the subject. Count time was 30 minutes. A lung phantom with 

variable chest wall thickness was developed for calibration of the new system. MDAs were listed as 

0.15 to 0.6 nCi for 241Am, 2.0 to 3.7 nCi for 234Th (assumed to be in equilibrium with 238U), and 0.17 to 

0.37 nCi for 235U, depending on a subject’s weight to height ratio. (Chest count MDAs are 

summarized in Table 5.3.2-1.) MDAs for direct measurement of 238Pu and 239Pu using the 17 keV xrays were calculated at times, but the values were extremely large relative 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 counts were probably not routine counts. MDAs for those 

counts were listed as 25 – 40 nCi and 0.5 – 1.5 μCi for 90Sr and 147Pm, respectively. A second chest 

counting system became operational in 1978. A phoswich detector became available and was used 

occasionally for special chest counts but was never implemented on a routine basis. 

Chest Counters
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A solid state germanium counting system using 3 planar HPGe detectors replaced the NaI detector in 

the Iron Room chair counter in 1983. 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). They also have a thin window on the end of the 

detector facing the chest for better transmission of low-energy photons. The detectors were 

positioned over the front of the chest (two over the right lung) with the subject in the supine position. 

Counting time was 2000 seconds. MDAs were quoted for “an average size person” as 0.1 nCi for 

241Am, 0.5 nCi 144Ce, 0.7 nCi of 234Th (238U), 0.05 nCi 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 plus was discernable by naked-eye inspection of the 

spectrum (Carbaugh et al 1988). 

Special chest counts, as follow-up to high routine chest counts or upon special request, were twice 

the normal counting time so the MDAs were about somewhat lower. 

The 3-detector system was soon (within about a year) upgraded to a 6-detector array, which allowed 

routine counting times to be reduced to 1000 seconds with nearly the same RDAs (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 6-detector array, this counter had only 4 detectors at first because of 

operational problems with the detectors. Counting times were increased to 2000 seconds for the 6-

detector array and 3000 seconds for the 4-detector array. 

In September 1994 the chest counter in the Stainless Steel Room was converted to a 4-detector array 

using larger area detectors. The same change was implemented in the Iron Room in June 1996. 

This configuration continues to the present. The routine counting time was increased to 3000 

Table 5.3.2-1. Routine chest counting detection 

levels. 

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 0.18c

U-238 1.8

U-235 0.12
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Period Radionuclide MDA (nCi)

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 chestwalls. 

Use highest value for default to cover large persons. 

b. Assumed MDA = (RDA)(4.65/3). Am-241 adjusted for 

95th percentile male chestwall (.2/.13)

c. Adjusted for 95th percentile male chestwall.

d. 144Ce and 154Eu no longer automatically reported for 

chest counts because now can be quantified in the Ge 

whole body counter.

e. Applies to the 6-detector array. Better sensitivity was 

obtained by the 4-large-area-detector array in the 

Stainless Steel Room. 

seconds for the larger area arrays in November 1995; special counts and recounts were 3600 

seconds. 

Ultrasonic measurements of chestwall thickness for workers that had activity in the lung began in 

about 1979 and continues today. So decision levels for non-detected activities use a weight-to-height 

ratio to estimate chestwall thickness, whereas detected activity is corrected for chestwall thickness 

using ultrasound. 

Individual-specific decision levels were reported to the database for each count, each radionuclide, 

starting in 1992. 

For in vivo counting, the assumption was made that 234Th was in equilibrium with 238U. This was a 

reasonable assumption at Hanford. Certainly, uranium recently separated from dissolved fuel was not 

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

depending on how long the material had taken 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 relative to chest counting. Chest counts were used to monitor for intakes of insoluble forms 

of uranium, which also were very old forms in terms of time since purification from decay progeny 

(e.g. machining on metal, uranium metallurgy studies). 

5.3.3

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

1956. (See also first part of chapter 5 for discussion on thyroid counting in 1945 and 46.) A letter to 

file, dated June 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 

Thyroid Counters 
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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 goes 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 also 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 right 

next to the neck. The MDA was given as 0.020 nCi for 131I for a 30-minute count. The exact same 

detector and MDA were included in a description of in vivo counting capabilities at the Whole Body 

Counting Facility in 1971 and again in 1985. 

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 uncovered yet. The 

MDA was listed as 0.11 nCi for a 1-minute count or 0.07 nCi for a 10-minute count, but there was no 

mention as to 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. The same counter is described for thyroid counting in 1982, except that the 

“reporting level” is given as 0.020 nCi; it’s 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 2000-second count. 

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

5.3.4

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 will show in the database 

almost always listed as special counts associated with known intakes. 

Head Counters and Other Counts

Since the mid 1980s, for intakes of plutonium or americium, head counts have been used to correct 

chest counts for activity in the bones of the chest region. Since the mid 1990s liver counts were 

added to the protocol for correcting chest counts to account for possible shine from the liver. 

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

sensitive plus there is the question as to what a head count means relative to the activity in the total 

skeleton. Hopefully the same worker will have 90Sr urinalysis results. The latter should be given 

preference as to confirming or quantifying an intake. 

5.3.5

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

decision levels until 1992. The decision level is listed under “limits” starting January 1992. 

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 may be available on the In Vivo Counter Results Form if it was sent to the worker’s 

personal radiation exposure history file. If not, assume the result of the count is100 nCi. 
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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 may or may 

not have been quantified. Additionally, the activity of one or more of the regular four radionuclides 

may have been marked as invalid because of overlap with the interfering peak or because of impact 

of the interfering peak on the spectrum stripping calculations. For the small activities involved, there 

is no merit in trying to recalculate or estimate actual quantities. It is claimant favorable to use the 

activities of 137Cs as given plus include the activity of the interfering radionuclide as given as well. 

Use 100 nCi for the interfering radionuclide if not given directly. 

Which radionuclides were 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, with 

only 137Cs being of interest to the dose reconstructor. 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. In 1995, with the start-up 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. During this time 

if a worker had a count using the coaxial HPGe whole body counter, up to 20 radionuclides may 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. 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. 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 chest counting were 241Am, 234Th, 

and 235U for anyone receiving a chest count. 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 that for plutonium workers, for instance, 

only the 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. Emphasis was 

on measuring exposure to radionuclides having the greatest impact relative to radiation protection 

standards (for instance, MPBB or CEDE), 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. Hence, 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, those data 

should be used 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. 
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5.4.1

Fission and activation product mixtures up through 1987 (when N Reactor shut down) were much 

more complex and variable. 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 strontium so the recorded result overestimated 

cerium and yttrium activity. The uncertainty associated with calibration or yield is overwhelmed by the 

uncertainty in guessing what the mixture was truly composed of and what the abundances of 

unmeasured radionuclides were. The relative abundances of radionuclides in a potential intake of 

mixed fission and activation products varied according to location (reactors, fuel separation facilities, 

waste management facilities), type of fuel, enrichment of fuel, amount of burn-up, and cooling time 

(i.e., time since removal of the fuel from the reactor). Reactor operators were most likely exposed to 

activation products, but contamination from leaking fuel rods, especially in fuel storage pools, cannot 

be ruled out. Even the exposure to activation products was considerably uneven and depended on 

which reactor components a worker had recently worked on, especially for radionuclides such as 51Cr 

or 46Sc. 

Fission and activation products

Due to complexity mentioned above, there is no straightforward way to determine the true makeup of 

an intake that resulted in a high fission product urinalysis result or to estimate how to apply the MDAs 

for missed dose calculations. 

It is, therefore, recommended not to use the fission product urinalysis results to determine 

intakes of fission or activation products. Therefore, for the years prior to implementation of whole 

body counting, default intakes based on air sampling are recommended. 

The default intakes are based on the same approach as was used for the period prior to the fission 

product urinalysis, namely chronic, 40-hour-per week exposure to the air concentration above which 

posting and respiratory protection was required. For the period 1948 through 1952, it is assumed that 

the requirement for respiratory protection at an air concentration of 1 x 10-9 μCi/cm3

 continued to 

apply. In 1953, if not earlier, the air concentration for 90Sr in the National Bureau of Standards 

Handbook 52 (NBS 1953), 2 x 10-10 μCi/ml, was used as the limiting concentration for exposure to 

particulate beta/gamma emitters. This value would have been used until approximately 1959 when 

limiting air concentrations were revised by NBS Handbook 69 (NBS 1959). 

For workers on the fission product bioassay program from 1948 through 1952, assume inhalation 

intakes of 6,600 pCi and ingestion intakes of 200 pCi per calendar day. For 1953 through the time 

when the worker was given a whole body count, assume inhalation intakes of 1,300 pCi and ingestion 

intakes of 40 pCi per calendar day. Use Table 5.7.2 to choose the radionuclide. These intakes are 

modes of a triangular distribution with a minimum of zero and a maximum of 2 times the mode. 

The recommendation above is not intended to preclude the DR from using fission product urinalysis 

results that are repeatedly above the reporting level or MDA. Hopefully, the worker’s file will provide 

some insight toward what radionuclide(s) the worker was exposed to. Problems with breakthrough of 

natural 40K were noted (Healy 1948), so a single high result might not be indicative of a true intake. 

The DR must keep in mind that the fission product urinalysis did not account for intakes of some 

fission products (e.g., 137Cs, 106Ru) and most activation products. 

If dose assessment based on fission product urinalysis is indicated, the following discussion may be 

helpful in interpreting the results. Note: the discussion may also be helpful in interpreting whole body 

count results for 1960 to present. 
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5.4.1.1 Reactor workers

Reactor workers would have been principally exposed to activation products; however, as yet it has 

not been determined how workers were monitored for possible intakes of activation products until 

1960, except by air sampling. Air samples were analyzed for total alpha and total beta so the mix of 

activation and fission products was not determined. The fission product urinalysis did not measure 

activation products. Because of once-through cooling water, effluents from the reactors show lots of 

short-lived radionuclides, but an interview of a health physicist (Marvin Smith) who worked at the early

reactors revealed that workers were not allowed into the rear face and rear tubing area, where shortlived activation products were present, for 12 hours after shutdown to allow for decay. So, workers 

were most likely exposed to the usual mix of activation products: 54Mn, 58Co, 60Co, 59Fe, 51Cr. The 

list of radionuclides detected in the first couple years of whole body counting also showed 46Sc, and 

99mTc (indicator for 99Mo). Another 100 Area health physicist (G. Yesberger) also said that the 

workers wore assault masks when working on contaminated parts of or equipment associated with the 

reactors. Both men said that airborne contamination was never an important consideration, dosewise, relative to the high external exposure rates in those areas. 

Based on what little information has been uncovered, if an intake of anything other than 131I, 24Na, or 

65Zn is indicated for a worker at the reactors, prior to 1960, the dose reconstructor should consider the 

possibility of intakes of 46Sc, 51Cr, 54Mn, 59Fe, 60Co, 99Mo as well. 

154Eu and 155Eu were activation products of concern at N Reactor because of activation of samarium 

balls used for neutron flux control. These radionuclides would have been included in the fission 

product urinalysis or would have been detected in whole body counts. If one is detected, then 

assume the intake included an equal amount of the other. 

After shutdown of N Reactor in 1987 the shorter lived activation products began to decay away. It 

would be safe to assume that only the europium isotopes and 60Co were available for intakes by 1989. 

Assume the following absorption types for reactor workers: Sc type S, Mn type F or M, Cr type F or S, 

Co type M or S, Sr type F, Mo type F or S, Cs type F, Eu type M, Fe type F or M. 

5.4.1.2 Separations Plants 

The following guidelines may be used to determine intakes from fission product urinalysis results if no 

other information is available about the radionuclide composition of the contamination. (See 

Attachment D 2.1 for basis for guidelines). 

(Assume the following absorption types: Ce type M, Y type M, Sr type F, Nb type M, Zr type M, Ru 

type F, Pm type M.)

Calculate the intake assuming the fission product activity is 50% 141Ce and 50% 144Ce. Then add 

intakes of the following (in multiples of the Ce intake): 

1948 - 1955

Bone: 1.0 91Y, 0.5 89Sr, 1.5 95Nb, and 0.5 103Ru. 

Liver: 1.5 95Nb, 1.0 91Y, and 0.5 103Ru. . 

GI: 1.5 95Nb, 1.0 95Zr, 1.0 91Y, and 0.5 103Ru.
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Lung: 1.5 95Nb, 1.0 95Zr, 1.0 91Y, 0.2 147Pm. . 

All other organs: 1.5 95Nb, 1.0 95Zr, 1.0 91Y, 0.2 147Pm.. 

Calculate the intake assuming the fission product activity is 144Ce. Add (in multiples of the 144Ce 

intake): 

1956 - 1960

Bone: 0.6 91Y, 0.4 89Sr, 0.8 95Nb, 0.1 90Sr, and 0.6 106Ru. 

Liver: 0.8 95Nb, 0.6 91Y, and 0.6 106Ru.

GI: 0.8 95Nb, 0.7 95Zr, 0.6 91Y, and 0.6 106Ru.

Lung: 0.8 95Nb, 0.7 95Zr, 0.6 91Y, 0.4 147Pm. 

All other organs: 0.8 95Nb, 0.7 95Zr, 0.6 91Y, 0.4 147Pm.

Because whole body counting became routine, whole body counts can be used to determine intakes 

of fission or activation products. If any of the main gamma-emitting fission products were detected 

and 90Sr urinalysis was not obtained, determine the intake of the gamma-emitting radionuclide, then 

add (in multiples of the gamma-emitting radionuclide intake) 1.0 90Sr, 7.0 89Sr, 12 91Y, and 4.0 147Pm. 

1961 - 1972

During the period when PUREX was shutdown, 1973 through 1983, considerable facility upgrades 

and maintenance activities were conducted; hence, exposures to contamination continued, but the 

mixture would not have contained much activity from short half-life radionuclides. Ratios from 2-yearcooled N Reactor 6% fuel would be conservative with the adjustment that 137Cs and 90Sr would have 

built up in contamination over the lifetime of the plant more than 144Ce or 106Ru. Consequently it would 

be claimant favorable but reasonable to assume equal intakes of the major five radionuclides left in 

the mix, i.e., if an intake of any one of these was incurred, then assume an equal intake of the others: 

144Ce, 137Cs, 147Pm, 106Ru, 90Sr. 

1973 - 1983

When PUREX restarted in November 1983, it mostly processed very long cooled fuel or blends of old 

fuel with some fuel cooled at least 180 days. Either way the short-lived beta-emitting contamination in 

the plant was not significant. If either of the cerium isotopes or ruthenium isotopes were detected, 

then because of the difference in MDAs, 137Cs should have also been detected. If not, it means that 

the cesium was reduced in the mixture due to processing activities and can be ignored. However, if 

90Sr urinalysis was not obtained in approximately the same time period, then it is claimant-favorable to 

assume (relative to the cerium or ruthenium intake) 0.6 90Sr and 1.0 147Pm. 

1984-1989

Because N Reactor shut down in 1987, fission product contamination at PUREX would have been 

almost entirely 137Cs, 90Sr and 147Pm in approximately equal activities. 

Post 1989

5.4.1.3 Waste Management Facilities (tank farms, evaporators, transfer lines)

By the time waste products reached tank farms and were further manipulated by the evaporators, B 

Plant, and settling in the tanks, ratios among fission products could be vastly different than in the fuel 
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going into the separations process. 144Ce, 106Ru, 137Cs and 241Am are usually associated with the 

supernatant and generally more available as contaminants. Plutonium, 90Sr, 147Pm are associated 

with the sludge. Lacking any other information, assume an intake of one radionuclide of the first 

group (supernatant or general contamination) exists in a mixture of the following: equal activities of 

144Ce, 106Ru, 137Cs, 0.1 99Tc, 0.1 90Sr, and 0.001 241Am. Assume an intake of one radionuclide of the 

second group (sludge) exists in a mixture of the following: equal activities of 90Sr and 147Pm, 0.1 137Cs, 

0.001 239Pu and 0.001 241Am. (Carbaugh, 1995) For absorption types, use either M or S for cerium, F 

or S for ruthenium, F for strontium, cesium and technetium, M for americium, and M or S for plutonium 

and promethium, depending on whether the organ of concern is associated with the respiratory tract 

or GI tract (S) or all other organs (F or M). 

5.5 INTERFERENCES, UNCERTAINTIES

5.5.1

Home collection of excreta samples started very early in the bioassay program; 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 is obtained the same day or within a few 

days that shows a dramatic decrease in activity or no detectable activity, then external contamination 

can be assumed. 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, the 

results should be used as recorded.

5.5.2

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

excreta measurements. These are listed in the database under Error and represent total propagated 

uncertainty (one σ) including counting uncertainty, yield uncertainty, and various other systematic 

uncertainties. These should be used when available. For excreta, uncertainty can also exist in the 

sample date. For routine samples an uncertainty of ± two weeks can be assumed. 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 ± two days is reasonable unless the sample is within 2-3 days of a 

known intake. 

Uncertainties

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

were 24-hour simulated samples (kit code 1), meaning the sample was collected over two eveningthrough-morning periods. Medley indicated that this sampling method produced only about half of a 

true 24-hour sample based on volume for a group of 9 workers over a 3-day period (Medley 1994); 

however, Hanford collection protocol was based on percent of day not volume so the true bias (when 

samples were collected according to procedure) was about 75% of a true 24-hour sample. If a worker 

has enough urine samples to establish the individual-specific excretion pattern, then a sample can be 

normalized to the individual’s expected 24-hour excretion. Generally, the error associated with 

collection time period results from under collection of a 24-hour volume. It is claimant favorable to 
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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-hour 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. 

For in vivo results, uncertainties were not reported until 1986 for detected radionuclides and 1993 for 

the default set of radionuclides. These were one sigma counting errors until 1995. Total propagated 

error was 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. Uncertainty associated with reproducibly 

positioning a person to get the same result was studied at Hanford and found to be about 5%. All 

calibrations are made using phantoms, and there is considerable uncertainty associated with the 

representativeness of phantoms versus humans. Just recently a study was done for whole body 

counting at Hanford using a 95th percentile reference man phantom. There was a low bias of about 

20% for the coaxial HPGe detector system for 662 and 1332 keV gamma rays. A similar value of 

uncertainty (±20%) can reasonably be assumed for the other whole body detectors (1-meter arc, 

shadow shield, and standup counters). 

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 

one-sigma uncertainty of about 20% for americium and uranium values in chest counting, not 

including correction for interferences from bone and liver. 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 likely range from 100% or more for levels 

near or below the MDA to 50% or more for activity above the MDA. 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, 

then a GSD of 5 uncertainty should be associated with an intake determined by chest counting. 

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

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

reporting level, is adequate except for chest counts for which 0.5 times the MDA should be used. 

If actual standard deviations or other indications of error are reported with a bioassay measurement 

result that exceed the above guidance, the reported value should be used. 

5.6 WORKERS WITH NO CONFIRMED INTAKES

5.6.1

Because plutonium, americium, and thorium stay in the body for a very long time, and because the 

urine excreta curve (activity per day excreted versus days after intake) has a small slope beyond the 

first year, an intake of plutonium or thorium that might have been missed in the 1950s or 60s because 

of poor detection capability, missed samples, or poor sampling after a potential intake, can still be 

confirmed by urinalysis obtained years later. This is especially true for type S materials, but even 

applicable to type M. For instance, the urine excretion curve for type M plutonium, thorium, and 

americium all decrease only a factor of 3-4 from one year to 4000 days (~11 years) after the intake. 

So if an intake is suspected but was not confirmed, the dose reconstructor can use the more sensitive 

Special Consideration for Plutonium, Americium, and Thorium
Effective Date: 11/24/2004 Revision No. 01 Document No. ORAUT-TKBS-0006-5 Page 47 of 57

urinalysis data obtained much later to determine a worst case intake. The MDA applicable at the later 

time can be used or, if there are many samples all showing no detection, then 0.5 times the MDA can 

be used for the urine value. 

5.6.2

Chronic intakes, or frequent, intermittent intakes that can be modeled as chronic, occurred for tritium 

and uranium. For other radionuclides, very low level, frequent, intermittent intakes may have occurred 

for the highest risk workers, such as operators and maintenance workers, at the reactors, separations

plants, and Plutonium Finishing Plant, but the intakes were below detectability at the time. For 

radionuclides with long residence times in the body, chronic intakes lead to a slow build-up of activity 

in the body and a concomitant increase in urinary excretion. For workers with many bioassay results 

over a long time but no confirmed intakes, a maximum chronic intake can be determined by using the 

MDA of the last sample as the upper bound of excretion assuming chronic intake for the entire 

exposure period. The MDA not the decision level should be used for this calculation. The rate of the 

chronic intake (pCi/day) needed to reach the MDA level of excretion varies with the duration of the 

intakes. A lower rate of intake is needed to reach the MDA level excretion if the duration of intakes is 

20 years as opposed to 2 years, for instance. 

Worst Case Chronic Intakes

Attachment D, Section D.3 provides tables of chronic intakes used to reach MDA levels of urinary 

excretion for plutonium, americium, and type S uranium. The tables are based on a unit MDA (1 

dpm/day). Adjustment of the actual chronic intake is linear with MDA, so if the true MDA is 0.02 

dpm/day, the actual intake is 0.02 times the table value. For the period when the plutonium MDA was 

for total alpha, Table 5.2.1-3 can be used to determine the isotopic composition of the intake. For the 

period when the plutonium MDA was for 239Pu directly, Table 5.2.1-3 can also be used to determine 

the other components of the mixture. (Do not use the 241Am chest count MDAs for worst case chronic 

intakes of plutonium.) Guidance on adjusting for different enrichments of uranium is given in Table 

5.2.5-1.

Attachment D Section D.3 also provides tables of chronic intakes used to reach MDA levels of urinary 

excretion or retained quantities in the whole body for radionuclides that have short-half lives or short 

retention in the body. For these radionuclides the daily urinary excretion or retained quantity in the 

whole body does not continue to increase throughout the exposure period; instead, equilibrium is 

reached quickly. For these radionuclides, if there were changes in the MDA throughout the period of 

employment, the calculation of daily intakes and cumulative intake should be made separately for 

each period; however, it is claimant-favorable and faster to just assume the highest MDA applies for 

the whole exposure period. 

For whole body counts, it is unreasonable to assume that a worker was exposed to all the 

radionuclides potentially reportable simply because an MDA was determined; on the other hand, for 

years Hanford only reported 24Na, 65Zn, 40K, and 137Cs, of which only 137Cs is of concern to dose 

reconstruction, while other fission and activation products were ignored if less than 1% of a MPBB. A 

recommended approach would be to use an indicator radionuclide to determine the intake, then add 

intakes of other radionuclides in the mixture, as discussed in Section 5.4. Since 1987 or 1988 the 

only whole body count radionuclides of potential exposure have been 137Cs and 154Eu (maybe 60Co 

into the early 1990s). 

 
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5.7 UNMONITORED WORKERS

From the start Hanford has always had a radiation protection program and many innovations in health 

physics were developed at Hanford (for instance, the first shadow shield whole body counter). 

Developing a bioassay program for plutonium and uranium was a major focus from the start; and atrisk workers were incorporated into the bioassay program as soon as possible. Air sampling 

programs, on the other hand, were used mostly to detect contamination spreads and to decide if an 

area had to be “on mask.” Air sampling results that were well below mask level (e.g, AEC Manual 

Chapter 0524 Table 1 concentrations) were generally ignored. Construction jobs were monitored if 

the work occurred in a contaminated facility and, supposedly, all outside jobs with potential for 

encountering underground contamination. But construction workers were neither consistently placed 

on routine bioassay nor consistently scheduled for termination bioassay. 

Due to the workplace monitoring program the probability that a worker received a large intake of 

radioactive material that was unmonitored and unnoticed is very remote (once bioassay programs 

were established). For instance, even among monitored workers, high routine bioassay results were 

rare except for tritium, uranium, and fission and activation products up through the 1970s. However, 

especially among construction workers, the probability of unmonitored, small intakes is larger. On the 

other hand, many workers had jobs that never required them to enter contaminated areas or to do so 

only rarely on tours or inspections. 

Under certain conditions, airborne effluents from one facility became air intakes for other facilities. 

Also workers were exposed to diluted effluents when walking between buildings or parking lots or 

while driving on the site. So, workers in buildings who did not enter contaminated or airborne areas 

and construction workers almost anywhere could have incurred environmental level intakes. 

Another point is that before 1994, workers with even a remote chance of exposure to workplace 

external radiation or workplace airborne contamination (as opposed to exposure from effluents) had a 

dosimeter. Through 1971 workers obtained film dosimeters before proceeding past badge houses at 

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

external dose had a minimal dosimeter, called the Hanford basic dosimeter. 

The missed 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. 

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.

For workers who wore a dosimeter but did not have any bioassay apply the following guidance. 

Hanford was a complex site that handled curie-to-megacurie quantities of many radionuclides, and the 

list of radionuclides is even longer when smaller, research and development levels are considered. 

For unmonitored workers, the only way to discern which radionuclides they might have encountered is 

to know what facilities they accessed. Such information is rarely included in a worker’s dosimetry file, 

especially if the worker was unmonitored. Nevertheless it is implausible that any unmonitored worker 

was exposed to all the radionuclides in use at Hanford, and assigning even small intakes based on 

bioassay MDAs for a majority of the possible radionuclides is unrealistic. 

Air sampling was performed in facilities from the earliest days of operation. Workers routinely 

exposed to areas with detectable airborne concentrations or performing high risk jobs that might 

produce airborne contamination were on routine bioassay schedules (except prior to about 1948). 

Historical criteria for placing workers on routine bioassay schedules have not been discovered for all 

times or all facilities and probably were facility-by-facility specific. However, all facilities had broad-
Effective Date: 11/24/2004 Revision No. 01 Document No. ORAUT-TKBS-0006-5 Page 49 of 57

scoped radiation protection programs that were designed to limit intakes of radioactive material and 

monitor workers at risk for intakes. Workers who had dosimeters but no routine bioassay would have 

been those who were only occasionally in areas with the potential for airborne contamination or with 

measured but low airborne contamination or had work locations distant enough from airborne 

contamination that the air was considered uncontaminated. For the period 1953 through 1988, entry 

into areas with airborne concentrations exceeding the maximum permissible concentrations (MPC) 

(also called derived air concentration, DAC, or concentration guide, CG, shown in Table 5.7-1) 

required respiratory protection. Typically workers entering such areas were on bioassay programs. 

From 1989 airborne areas were controlled at 10% of the DAC (DOE Order 5480.11 Most regularly 

occupied areas had no detectable airborne radioactivity contamination. Also workers were placed on 

routine bioassay programs if they had the potential to incur intakes at 10% of the quarterly dose limit 

from 1968 through 1988 (AEC Manual Chapter 0524) and 2% of the annual limit from 1989 to present 

(DOE Order 5480.11). When worker bioassay requirements were established by radiation work 

permits (circa late 1980s), the requirements were generally set lower than the 2% DOE limit to 

compensate for the possibility of working under several radiation work permits in a year. 

Assumptions for intakes by workers who wore dosimeters but had no bioassay [except perhaps 

baseline or termination bioassay] (e.g., administrative workers, support engineers, design engineers, 

most construction workers, quality assurance engineers, security guards) depend on the time period 

of exposure and the facility or general category of facility they accessed.

For the earliest years before bioassay programs were well established, use 1.0 times the guidance in 

Section 5.1 for workers likely to have been in work areas several times a week, and 0.1 times those 

values for workers in job categories not expected to have come into direct contact with the radioactive 

materials and to have only occasionally entered contaminated work places. If the DR does not have 

sufficient information to distinguish the frequency, use 1.0 times the guidance in Section 5.1. If the 

DR cannot discern the appropriate facility, use the most claimant-favorable facility. 

For the years following the periods cited in Section 5.1, the DR should base the intake assessments 

on the daily intake values presented in Table 5.7.1. Those values were based on a work time 

breathing rate of 1.2 m3

/hr for the hours/week discussed below divided by 7 days/week to produce a 

daily chronic intake rate. (Use of 7 days a week is required for the computer code.)

For the years 1947 through 1952, the daily intakes were based on the assumption that exposure to 

airborne concentrations occurred at just under the “respiratory protection required” value (Patterson 

1949) for 4 hours each week or 10% of the “respiratory protection required” value for 40 hours each 

week. This assumes that these workers either 1) occasionally entered areas with airborne 

contamination just below requirements for posting and respiratory protection or 2) were located full 

time in an area with low but arguably not zero airborne contamination. The Patterson letter did not list 

a “respiratory protection required” concentration for 131I. Monitoring for 131I uptake was performed by 

frequent thyroid scans; for instance, at the fuel separation facilities typically 500-600 thyroid scans 

were done each month. Use 0.1 times the vapor intake value used for the pre-1947 period, i.e., 7.5 x 

104 pCi/d.

For 1953 through 1988, the daily intakes were based on the assumption that exposure to airborne

contamination at 10% of the limiting air concentration occurred for 4 hours each week. For particulate 

Table 5.7-1. Intakes for workers with external monitoring but essentially no bioassay. 

Period Radioactive material

Limiting air 

concentration 

(μCi/cm3)

a

Daily intake 

(pCi/d)
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Period Radioactive material

Limiting air 

concentration 

(μCi/cm3)

a

Daily intake 

(pCi/d)

Start to 1946b Use values in Section 5.1 for workers exposed several days a week 

and 0.1 times those values for workers exposed only occasionally. 

1947-52c Particulate beta/gamma emitters 1 x 10-9 690

Particulate alpha emitters 2 x 10-12 1.4

Uranium (soluble or insoluble 5 x 10-11 34

I-131 (vapor) 7.5 x 104

Tritium (HTO) 1 x 10-5 1.0 x 107d

1953-67 Particulate beta/gamma emitters 3 x 10-10 21

Particulate alpha emitters 2 x 10-12 0.14

U nat. insoluble 6 x 10-11 4.1

U nat. soluble 7 x 10-11 4.8

I-131(vapor) 9 x 10-9 620

Tritium (HTO) 2 x 10-5 2.1 x 106d

1968-88 Particulate beta/gamma emitters 1 x 10-9 69

Particulate alpha emitters 2 x 10-12 0.14

U-238 insoluble 6 x 10-10 41

U-238 soluble 7 x 10-11 4.8

I-131 9 x 10-9 620

Tritium (HTO) 5 x 10-6 5.1 x 105d

1989-present Particulate beta/gamma emitters 2 x 10-9 69

Particulate alpha emitters 2 x 10-12 0.069

U-234 insoluble (class Y) 2 x 10-11 0.69

U-234 soluble (class D) 5 x 10-10 17

I-131 2 x 10-8 690

Tritium 2 x 10-5 1.0 x 106d

a. The AEC, ERDA and DOE terms used for limiting air concentrations include concentration 

guide, maximum permissible concentration and derived air concentration. 

b. Approximate date; see Section 5.1 for facility-specific dates.

c. Predates formal terminology; values given are when respiratory protection was required 

(Patterson 1949).

d. Total intake including inhalation and skin absorption.

sampling, the limiting air concentration used to interpret total beta counts on air filters was based on 

90Sr, and the limiting air concentration for total alpha counts was based on 239Pu. The latter stayed 

constant throughout the period at 2 x 10-12 μCi/cm3

; but the limiting concentration for 90Sr varied from 

3 x 10-10 μCi/cm3 in ICRP 2 (1959) and the National Bureau of Standards Handbook 69 (1959) to 1 

x10-9 μCi/cm3 in AEC Manual Chapter 0524 and 2 x 10-9 μCi/cm3 in DOE Order 5480.11 (implemented 

1989). In the uranium-only facilities the uranium limiting air concentration was applied to total alpha 

counts.

For 1989 through present, because of stricter regulations and radiation area entry control 

mechanisms, the approach was based on exposure at 5% of the limiting air concentration for 4 hours 

each week.

Chronic intake is assumed for all these exposures. The assumed daily intake amounts for key 

radionuclides or radionuclide categories are listed in Table 5.7-1.

The other aspect to consider is how these intakes relate to facilities. That guidance is given below. 

Where two or more choices of radionuclides or absorption types are given, the DR should use the 

most claimant-favorable choice.
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100 Area Reactors 

• While operating, 1944-87

Table 5.7-2. Beta/gamma radionuclide chooser for unmonitored workers. 

: Use particulate beta/gamma and 131I intakes. Activation products 

were the principal sources of contamination; however, exposure to fission products cannot be 

disregarded; see Table 5.7-2 to determine the radionuclide that maximizes the dose to the 

organ of concern.

Organ of concern

Beta/gamma radionuclide and absorption type

Operating reactors

Fuel separation plants, 

waste management 

facilities

Adrenals Co-60 S Ru-106 F

Bone surface Sr-90 F Sr-90 F

Brain Ru-106 F Ru-106 F

Breast Ru-106 F Ru-106 F

Colon Ru-106 S Ru-106 S

Esophagus Ru-106 F Ru-106 F

Extra thoracic Co-60 S Ru-106 S

Gall bladder Ru-106 F Ru-106 F

Heart Co-60 S Ru-106 F

Kidneys Ru-106 F Ru-106 F

Liver Ce-144 M Ce-144 M

Lower large intestine Ru-106 S Ru-106 S

Lung Ru-106 S Ru-106 S

Muscle Ru-106 F Ru-106 F

Ovaries Ru-106 F Ru-106 F

Pancreas Ru-106 F Ru-106 F

Red bone marrow Sr-90 F Sr-90 F

Small intestine Ru-106 F Ru-106 F

Skin Ru-106 F Ru-106 F

Spleen Ru-106 F Ru-106 F

Stomach Ru-106 F Ru-106 F

Testes Ru-106 F Ru-106 F

Thymus Co-60 S Ru-106 F

Thyroid Ru-106 F Ru-106 F

Upper large intestine Ru-106 F Ru-106 F

Urinary bladder Ru-106 F Ru-106 F

Uterus Ru-106 F Ru-106 F

a. Use Eu-154 M instead if individual worked at N Reactor.

• When not operating, 1988 to present

300 Area Test Reactors, 1944-78: Use particulate beta/gamma and tritium intakes. See Table 5.7-2 

to determine the radionuclide that maximizes the dose to the organ of concern.

: Use particulate beta/gamma only, assign to 137Cs (F), 

90Sr (F), or 60Co (M or S).

400 Area, 1980-94: Use particulate beta/gamma intake with 24Na (F) as the radionuclide.

200 Area Fuel Separation Facilities

• T-Plant, 224-T, 1945-56; B-Plant, 224-B, 1945-52; Redox, 224-S, 1952-67; Purex 1956-92

•

: 

Use particulate beta/gamma, particulate alpha, and 131I intakes. Use Table 5.7-2 to determine 

the beta/gamma radionuclide intake. Input the alpha intake as either 234

U (F) or 239Pu (M or S) 

B-Plant, 225-B, 1953-present: Use particulate beta/gamma intake assuming either 137Cs (F) or 

90Sr (F). 
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Plutonium Finishing Plants (various Z buildings and 233-S), 1945 to present: Use particulate 

alpha as 239Pu (M or S).

221-U, 224-U (UO3 Plant), 1952-1993: Use soluble uranium intake. Input as 234U (F). 

108-B Tritium Extraction Facility, 1949-64: Use tritium intake.

C-Plant (Hot Semiworks)

• 1949-56

•

: Use particulate beta and particulate alpha intakes. Use Table 5.7-2 for beta/gamma 

radionuclide intake (fuel separation). Input the alpha intake as either 234U (F) or 239Pu (M or S). 

1957-67

300 Area Fuel Fabrication Facilities (303, 306, 313, 314, 333), 1944-88: Use insoluble uranium 

intake. Input as 234U (M or S).

: Use particulate beta intake. Assign to either 137Cs (F) or 90Sr (F).

209-E and 120 Critical Mass Labs, 1950-86: Use particulate alpha intake as 239Pu (M or S)

308 Plutonium Fuels Pilot Plant, 1960-1990: Use particulate beta and particulate alpha intakes. 

Use Table 5.7-2 for beta/gamma intake (reactor) and 239Pu (M or S) for alpha intake.

324 Building, 1966 to 2001: Use particulate beta and particulate alpha intakes. Use 137Cs (F) or 90Sr 

(F) for beta intake and 239Pu (M or S) for alpha intake.

325 Building, 1953 to present: Use particulate beta and particulate alpha intakes. Use Table 5.7-2 

for beta/gamma intake (waste management) and 239Pu (M or S) for alpha intake.

327 Building, 1953 to 1987: Use particulate beta/gamma and particulate alpha intakes. Use Table 

5.7-2 for beta intake (reactor). Input the alpha intake as either 234U (F) or 239Pu (M or S) 

Tank Farms and Evaporators, 1944 to present: Use particulate beta intake and 0.5 times the alpha 

intake (to account for a small alpha component in the waste). Use Table 5.7-2 for beta/gamma intake 

(waste management) and 239Pu (M or S) for alpha intake through 1987. Use 137Cs (F) or 90Sr (F) for 

the beta intake after 1987.

Unknown or highly variable work locations: Use one of the following combinations that is most 

claimant favorable:

• particulate beta intake using Table 5.7-2 plus 131I intake and tritium intake 

• particulate beta intake using Table 5.7-2 and particulate alpha intake as 239Pu (M or S)

• either soluble or insoluble uranium intake using 234U type F for the soluble intake and 234U type 

M or S for the insoluble intake 
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Effective Date: 11/24/2004 Revision No. 01 Document No. ORAUT-TKBS-0006-5 Page 57 of 57

GLOSSARY

active

A term used in the 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 cause by the tritium beta particle.”

aging

In the context of reactor fuel and mixtures of plutonium isotopes, aging refers to the time since 

241Am was separated from the plutonium mixture.

cooling

In the context of reactor fuel, cooling refers to 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, means 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 2 

consecutive nights to simulate a 24-hour sample or 4 consecutive nights to simulate a 48-hour 

sample.