ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document for the Hanford Site -- Occupational 

Internal Dose

Document Number:

ORAUT-TKBS-0006-5 

Effective Date: 10/15/2003

Revision No.: 00 

Controlled Copy No.: ________

Page 1 of 46

Subject Expert: Don Bihl

Approval: ___Signature on File__________ Date: 10/15/2003

Edward D. Scalsky, Document Owner

Approval: ___Signature on File__________ Date: 10/15/2003

Judson L. Kenoyer, Task 3 Manager

Concurrence: _Signature on File_________ Date: 10/19/2003

Richard E. Toohey, Project Director

Approval: __Signature on File___________ Date: 10/16/2003

James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

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

Page

3

Record of Issue/Revisions ...................................................................................................................3

Acronyms and Abbreviations ...............................................................................................................4

5.1 Occupational Internal Dose.........................................................................................................5

5.2 In Vitro Minimum Detectable Activities, Analytical Methods, and Reporting Protocols............... 10

5.2.1 Plutonium.....................................................................................................................11

5.2.2 Americium....................................................................................................................16

5.2.3 Curium .........................................................................................................................18

5.2.4 Tritium..........................................................................................................................18

5.2.5 Uranium .......................................................................................................................20

5.2.6 Fission Product Analysis .............................................................................................. 24

5.2.7 Strontium .....................................................................................................................25

5.2.8 Promethium .................................................................................................................26

5.2.9 Polonium......................................................................................................................27

5.2.10 Neptunium ...................................................................................................................28

5.2.11 Other Limited-Exposure Radionuclides ........................................................................ 28

5.3 In Vivo Minimum Detectable Activities, Analytical Methods, and Reporting Protocols ............... 29

5.3.1 Whole Body Counters .................................................................................................. 29

5.3.2 Chest Counters............................................................................................................ 33

5.3.3 Thyroid Counters ......................................................................................................... 35

5.3.4 Head Counters and Other Counts................................................................................ 36

5.3.5 General Notes about Items in the Database................................................................. 36
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5.4 Mixtures ....................................................................................................................................37

5.4.1 Reactor workers:.......................................................................................................... 38

5.4.2 Separations Plants....................................................................................................... 38

5.4.3 Waste Management Facilities (tank farms, evaporators, transfer lines)........................ 40

5.5 Interferences, Uncertainties ...................................................................................................... 40

5.5.1 Contamination of Samples ........................................................................................... 40

5.5.2 Uncertainties................................................................................................................ 40

5.6 Workers with No Confirmed Intakes.......................................................................................... 42

5.6.1 Special Consideration for Plutonium, Americium, and Thorium.................................... 42

5.6.2 Worst Case Chronic Intakes ........................................................................................ 42

5.7 Unmonitored Workers ...............................................................................................................43

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

Glossary ............................................................................................................................................46

LIST OF TABLES

Table

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

Page

5

5.1-2 Codes and radionuclides associated with bioassay at Hanford............................................ 7

5.2.1-1 Routine plutonium urinalysis detection levels..................................................................... 12

5.2.1-2 MDAs for nonroutine Pu excreta analyses ......................................................................... 13

5.2.1-3 Activity composition of Hanford reference weapons-grade plutonium mixture.................... 14

5.2.1-4 Activity composition of Hanford reference fuel-grade plutonium mixture ............................ 15

5.2.1-5 Activity composition of Hanford reference commercial power fuel-grade 

plutonium mixture .............................................................................................................. 15

5.2.2-1 Routine 241Am urinalysis detection levels ........................................................................... 17

5.2.2-2 MDAs for nonroutine 241Am excreta analyses .................................................................... 17

5.2.3-1 Routine Cm urinalysis detection levels............................................................................... 18

5.2.3-2 MDAs for nonroutine Cm excreta analyses........................................................................ 18

5.2.4-1 Routine tritium urinalysis detection levels .......................................................................... 19

5.2.5-1 Radiological characteristics of Hanford uranium mixtures .................................................. 20

5.2.5-2 Impurities in recycled uranium at Hanford.......................................................................... 21

5.2.4-3 Inhalation class for Hanford uranium compounds .............................................................. 21

5.2.5-4 Routine uranium urinalysis detection levels ....................................................................... 23

5.2.6-1 Routine fission product urinalysis detection levels ............................................................. 24

5.2.7-1 Routine 90Sr urinalysis detection levels............................................................................. 26

5.2.8-1 Routine 147Pm urinalysis detection levels......................................................................... 27

5.2.9-1 Routine 210Po urinalysis detection levels ............................................................................ 28

5.2.9-2 MDAs for nonroutine 210Po excreta analyses ..................................................................... 28

5.3.1-1 Routine whole body counting detection levels.................................................................... 31

5.3.1-2 Mean body burdens of 137Cs from fallout in the United States............................................ 33

5.3.2-1 Routine chest counting detection levels ............................................................................. 34
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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.
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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

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. 

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. 
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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. 

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. 

Reactor workers, 1944-49: Assume intakes of 200 nCi per year each of 46Sc, 51Cr, 54Mn, 59Fe, 60Co, 

90Sr, and 137Cs. (See Section 5.4.1 for absorption types.) (These values differ from assumed values 

for later years, Section 5.4.1, reflecting improved radiation protection programs over the years.) 

Separations plants, 231-Z, 1944-46: Assume intakes of plutonium alpha (6%Pu mix, see Table 

5.2.1-3) absorption class M at 8x10-13 μCi/cm3 for 7 hr per day, 5 days per week. Assuming a 

breathing rate of 1.5 m3

/hr, this translates to 6 pCi per day chronic intake. This chronic intake 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). Assume absorption type M. Also assume chronic intake of 131I of 

1.3x10+6 pCi per day (type F, 5 μm AMAD), which will produce equilibrium thyroid burdens at just 

under the 2-μCi tolerance level. [When IMBA can handle vapors, the better intake value is 7.5 x 10+5

pCi per day assuming a type F SR-1 vapor (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 will be able to differentiate these 

workers from general workers at the separations plants. 

300 Area uranium fabrication buildings (313, 314), 1944-47: Assume chronic intakes of 1500 pCi 

per day natural uranium (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. 

Laundry 1944-46: Assume chronic intakes of plutonium alpha (6% Pu mix, absorption class M) of 34

pCi per day based on an air concentration of 2 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 
Effective Date: 10/15/2003 Revision No. 00 Document No. ORAUT-TKBS-0006-5 Page 7 of 46

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

methods, routine urine sampling and analysis for plutonium 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
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Code Description Comment

BR 82 Bromine 82

C 14 Carbon 14

CE141 Cerium 141

CE143 Cerium 143

CE144 Cerium 144

CF249 Californium 249

CM242 Curium 242

CM244 Curium 244

CO 58 Cobalt 58

CO 60 Cobalt 60

CR 51 Chromium 51

CS134 Cesium 134

CS137 Cesium 137

EU152 Europium 152

EU154 Europium 154

EU155 Europium 155

EU156 Europium 156

EV155 ? Probably a typographical error for Eu-155 that got left in the 

database 

EV156 ? Probably a typographical error for Eu-156 that got left in the 

database 

FE 59 Iron 59

FP Fission products

GA Gross alpha

GB Gross beta

GELI Gamma-GeLi detector Excreta scheduling code for a gamma scan with a germanium 

detector

GOK 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

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

IRA Radium isotopic Excreta scheduling code

IR192 Iridium 192 Excreta scheduling code

ISCP Sequential Sr 90 Ce Pm Excreta scheduling code

ISPEC Gamma spectroscopy Excreta scheduling code

ISR Sr isotopic Excreta scheduling code

ITH Thorium isotopic Excreta scheduling code

ITPAC Seq isotopic Pu, Cm & 

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

NP239 Neptunium 239

PB210 Lead 210

PB212 Lead 212

PM147 Promethium 147

PO210 Polonium 210

PR144 Praseodymium 144

PU Plutonium alpha Total alpha from Pu isotopes after separation

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

PU238 Plutonium 238

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

PU240 Plutonium 240

PU241 Plutonium 241

PU242 Plutonium 242

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

QUS 1 U Same as QUS

QUS 2 U Same as QUS

RA224 Radium 224

RA225 Radium 225

RA226 Radium 226

RA228 Radium 228

RH106 Rhodium 106

RND Radon daughters

RU103 Ruthenium 103

RU106 Ruthenium 106

S 35 Sulfur 35

SB124 Antimony 124

SB125 Antimony 125

SCP Sequential Sr-total Ce Pm Excreta scheduling code

SM153 Samarium 153

SR Strontium Total radiostrontium by beta counting

SR 89 Strontium 89

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

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

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 

• 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.

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 
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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. 

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 

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

This practice [of recording lower detection levels] was discontinued and the more conservative 0.05 

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

Table 5.2.1-1. Routine plutonium urinalysis detection levels. 

Period

MDA, 

dpm/ sample

Decision Level, 

dpm/sample

Measured Quantity

Prior to June 1949 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 0.027 “

9/55 to 9/55 0.04 “

10/55 to 9/30/83 0.05a 0.025a “

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 TPUb “

9/2001 to present 0.02 2 x TPUb “

a. 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.

b. Xb is mean of blanks and TPU is total propagated uncertainty. 

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

included 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. 

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. 
Effective Date: 10/15/2003 Revision No. 00 Document No. ORAUT-TKBS-0006-5 Page 13 of 46

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 (c) 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 

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, these fecal samples should be interpreted as chronic intakes, 

not as acute intakes,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 
Effective Date: 10/15/2003 Revision No. 00 Document No. ORAUT-TKBS-0006-5 Page 14 of 46

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 239Pu 

to 238Pu ratio implies weapons grade 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. 

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 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 1.20 1.26 1.30 1.34 1.36 1.38

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

Pu alpha:241Am NA 14.6 8.63 6.67 5.72 5.18 4.84 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 
Effective Date: 10/15/2003 Revision No. 00 Document No. ORAUT-TKBS-0006-5 Page 15 of 46

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.

Table 5.2.1-4. Activity composition of Hanford reference 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-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 3.87 3.05 2.40 1.88 1.48 1.17 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.

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

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

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% 
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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 

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 is most claimant favorable, i.e., 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 inhalation 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. The first 

documentation found so far is a memo to file from John J. Jech, Senior Development Engineer in the 

Personnel Dosimetry Services, dated September 1969, that 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’s personal recollection provided some confirmation that this might have 

been the earliest date for americium urinalyses. The procedure was described 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 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 
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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. 

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

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 

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.

plutonium mixture. There is evidence of a few intakes of pure 241Am prior to 1969, involving usual 

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 
Effective Date: 10/15/2003 Revision No. 00 Document No. ORAUT-TKBS-0006-5 Page 18 of 46

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. 

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 4.2.3-1 and non-routine excreta analyses are provided in Table 4.2.3-2.

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.

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 

Tritium
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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 

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 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 1961

was based on earliest reference to liquid scintillation counting.
Effective Date: 10/15/2003 Revision No. 00 Document No. ORAUT-TKBS-0006-5 Page 20 of 46

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 

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 

about 1958) and recycled uranium starting in 1957 (fuel fabrication shops) or 1958. 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. 

Uranium 

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

234

U 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.
Effective Date: 10/15/2003 Revision No. 00 Document No. ORAUT-TKBS-0006-5 Page 21 of 46

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.

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 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 

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 claimant-favorable 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 5.2.4-3. 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 a Monday–morning-only sampling 

circa early 1980s. Change over should be clear in the records. 
Effective Date: 10/15/2003 Revision No. 00 Document No. ORAUT-TKBS-0006-5 Page 22 of 46

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

c. 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). [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 wet-ashing) with methyl isobutyl ketone and ammonium hydroxide. The detection 

limit for the latter was listed as 0.05 μg/L but the recoveries were about the same for both methods 

so the latter must have used a 10 times larger aliquot. Based on requirements in later statements 

of work, it is assumed that the first method was used for routine analyses. 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. 
Effective Date: 10/15/2003 Revision No. 00 Document No. ORAUT-TKBS-0006-5 Page 23 of 46

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 0.15/0.10(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.

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 

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

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). 
Effective Date: 10/15/2003 Revision No. 00 Document No. ORAUT-TKBS-0006-5 Page 24 of 46

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

Table 5.2.6-1. Routine fission product 

urinalysis detection levels. 

Period MDA 

1948 to 2/1956 30 dpm/sample 

3/1956 to 10/1958 70 dpm/samplea

11/1958 to 1960s 31 dpm/sample

1970b 5 dpm/sample

a. Recorded as 3.1 or 3.17 E-5 μCi/ sample; 

not clear if this was intended to be the MDA 

or just a reporting level.

b. Listed in the bioassay contract but probably 

not used; replaced by whole body counting 

and 90Sr urinalyses.

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, 
Effective Date: 10/15/2003 Revision No. 00 Document No. ORAUT-TKBS-0006-5 Page 25 of 46

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 

abundances of all the fission products, relative to each other, varied (considerably) as a function of 

the time from when the reactor fuel was removed from the core and allowed to cool to when the 

contamination was inhaled or ingested. 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 

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. 

Strontium

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.
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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 or1963 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.

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.

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

Period MDA or MCA

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

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. 

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 

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 

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. 
Effective Date: 10/15/2003 Revision No. 00 Document No. ORAUT-TKBS-0006-5 Page 28 of 46

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. 

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 

throughout the history of Hanford. The following discussion, addressing 14C, 232Th, radon, 90Y, 227Th, 

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

Other Limited-Exposure Radionuclides

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 
Effective Date: 10/15/2003 Revision No. 00 Document No. ORAUT-TKBS-0006-5 Page 29 of 46

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. 

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 

Whole Body Counters
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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. 

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, 95Zn, 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 

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 

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). 
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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

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

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. 
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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. 

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

reconstruction, only the 137Cs is of interest because 40K is strictly a natural source, and the 65Zn and 

24Na came from drinking water, the 65Zn from water in the cities surrounding Hanford and the 24Na 

from drinking water at the reactors. 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

From NCRP Report No. 94.

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 x-rays 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

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). 
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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

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. 

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 0.7 times lower. 
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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 

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 

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. 

Thyroid Counters 

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 
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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. 

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 
Effective Date: 10/15/2003 Revision No. 00 Document No. ORAUT-TKBS-0006-5 Page 37 of 46

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. 

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 the uncertainty 

associated with calibration 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. 

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. However, many of the principal 

fission products target the same organs. For instance, of the top 20 fission products (by activity) 

produced in Hanford reactors from 1945-60, isotopes of Sr, Y, Nb, Zr, Ce, La, Sb, Sn, Pm, Pr, Te, Nd, 
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Eu are bone seekers; Y, Nb, Ce, Pm, Pr, Nd, Eu are also liver seekers; Zr, La, Sb, Sn also distribute a 

significant component to “all other tissues,” Cs and Ru distribute evenly to all tissues except, because 

of small uptake from the gut, Ru delivers more dose to the GI tract organs; Ba principally affects the 

GI tract; and I concentrates in the thyroid. 

5.4.1

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 who worked at the early reactors 

(Marvin Smith) revealed that workers were not allowed into the rear face and rear tubing area, where 

short-lived 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. 

Reactor workers: 

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 assume 

intakes of 46Sc, 51Cr, 54Mn, 59Fe, 60Co, 99Mo occurred as well. The amounts of such intakes would not 

necessarily correlate well to any fission product radionuclides. Instead about 100,000 pCi (intake) of 

each would be a conservative estimate, which is roughly 1% of the MPBB of each. 

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. 

5.4.2

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 3.2 for basis for guidelines). 

Separations Plants 

(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.)

Bone: Calculate the intake assuming the fission product activity is 141Ce. Then add intakes of the 

following (in multiples of the 141Ce intake): 1.0 91Y, 0.5 89Sr, 1.5 95Nb, and 0.5 103Ru. Calculate the 

dose assuming that the cerium intake is 144Ce. 

1944 - 1955

Liver: Calculate the intake assuming the fission product activity is 141Ce. Then add intakes of the 

following (in multiples of the 141Ce intake: 1.5 95Nb, 1.0 91Y, and 0.5 103Ru. Calculate the dose 

assuming that intake is 144Ce. 
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GI: Calculate the intake assuming the fission product activity is 141Ce. Add intakes (in multiples of the 

141Ce intake): 1.5 95Nb, 1.0 95Zr, 1.0 91Y, and 0.5 103Ru.

Lung: Calculate the intake assuming the fission product activity is 141Ce. Add 1.5 95Nb, 1.0 95Zr, 1.0 

91Y, 0.2 147Pm. Calculate the dose assuming that the intake is 144Ce. 

All other organs: Calculate the intake assuming the fission product activity is 141Ce. Add intakes of 1.5 

95Nb, 1.0 95Zr, 1.0 91Y, 0.2 147Pm. Calculate the dose assuming that the intake is 144Ce. 

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

144Ce intake): 0.6 91Y, 0.4 89Sr, 0.8 95Nb, 0.1 90Sr, and 0.6 106Ru. 

1956 - 1960

Liver: Calculate the intake assuming the fission product activity is 144Ce. Add: 0.8 95Nb, 0.6 91Y, and 

0.6 106Ru.

GI: Calculate the intake assuming the fission product activity is 144Ce. Add: 0.8 95Nb, 0.7 95Zr, 0.6 91Y, 

and 0.6 106Ru.

Lung: Calculate the intake assuming the fission product activity is 144Ce. Add 0.8 95Nb, 0.7 95Zr, 0.6 

91Y, 0.4 147Pm. 

All other organs: Calculate the intake assuming the fission product activity is 144Ce. Add 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
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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

5.4.3

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 

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). 

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

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. If the dataset shows an unusually high urinalysis result for a radionuclide other than tritium or 

uranium, and if follow-up samples were collected that were not consistent with the high result, then 

the high result may be considered an outlier. However, if the result is not obviously an outlier, then it 

is claimant-favorable to assume the result is real. For plutonium, if a single high result exists in the 

database and later (even years later) bioassay results at lower MDAs exist that do not show detection, 

it is reasonable to assume the high result was an outlier. 

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 

Uncertainties
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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. 

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 

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. 

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.

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

Implementation Guide (NIOSH 2002b) is adequate and should be used, namely the standard 

deviation is 0.3 times the MDA or reporting level, except for chest chests for which 0.5 times the MDA 

should be used. For results greater than 3 times the MDA or reporting level, the standard deviation 

can be assumed to be 0.1 times the result, based on Currie’s quantification level (Currie 1968). 

Actual tests for in vivo counts of phantoms show even smaller uncertainty, but 0.1 is good for broad 

applications. If actual standard deviations or other indications of error are reported with a bioassay 

measurement result, the reported value should be used. 

For intake estimates during the early periods prior to implementation of routine bioassay programs, 

where intakes are based on scanty air concentration data, a geometric uncertainty of x/÷ 2 is 

reasonable. 
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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 or otherwise 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 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. 

Special Consideration for Plutonium, Americium, and Thorium

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. 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 must be made separately for each 

period. Or 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 
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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). 

Using the mixtures in Section 5.4 does not work well when the worker’s work location is unknown or 

was variable. In that case it is claimant-favorable to pick the fission or activation product that 

produces the highest dose to the organ of interest rather than using an indicator radionuclide. To 

assist with this determination, a spreadsheet Radionuclide Chooser that produces 50-year committed 

doses to various organs from chronic intakes producing MDA-level body burdens has been developed 

and is recommended (available from the MJW Task 5 manager). 

5.7 UNMONITORED WORKERS

From the start Hanford has always had a strong radiation protection program and many innovations in 

health physics were developed at Hanford (for instance, the first shadow shield whole body counter). 

Developing a robust bioassay program for plutonium and uranium was a major focus right from the 

start; and at-risk 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 (MPC 

Table 1 values) 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 rigorous workplace monitoring 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. 

One other applicable point is that up to 1992, workers with even a remote chance of being exposed to 

external radiation had a dosimeter, in most cases, this was a single-chip TLD, referred to as the 

Hanford basic dosimeter.

These considerations lead to the following reasonable, yet claimant-favorable, assumptions for 

unmonitored workers at Hanford.

• If a worker’s record shows no bioassay and no evidence of ever having worn a dosimeter, the 

unrecorded internal dose should be based on environmental intake only.
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• If a worker wore a dosimeter, then the unrecorded internal dose would be no greater than that 

for a worker who was monitored but had no bioassay results exceeding reporting levels. 

In the latter case, use the guidance in section 5.6.2 in conjunction with period-specific MDAs levels 

and facility-specific radionuclides of concern to estimate an upper bound of possible unrecorded 

intakes. Claimant interview information may be helpful in determining which facilities the person 

worked at or near. 
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REFERENCES
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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. 

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.