ORAU Team

Dose Reconstruction Project for NIOSH

Technical Basis Document for the Argonne National 

Laboratory – West – Occupational Internal Dose

Document Number:

ORAUT-TKBS-0026-5 

Effective Date: 01/20/2005 

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 26

Subject Expert: Norman D. Rohrig

Document Owner

Approval: Signature on File Date: 01/20/2005

Norman D. Rohrig, TBD Team Leader

Approval: Signature on File Date: 01/20/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 01/20/2005

Richard E. Toohey, Project Director

Approval: Signature on File Date: 01/20/2005

James W. Neton, Associate Director for Science

Supersedes:

None

TABLE OF CONTENTS

Section 

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

Page

3

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

5.1 Introduction and Historical Overview...........................................................................................6

5.1.1 Argonne National Laboratory-West ....................................................................................7

5.1.2 Radioactive Nuclides of Concern and Solubility.................................................................. 7

5.2 Radiological Protection Program Management and Support ....................................................... 8

5.2.1 Bioassay Programs ............................................................................................................8

5.2.2 Internal Dose Records........................................................................................................8

5.2.3 Data Codes and Investigation Levels .................................................................................9

5.3 Internal Dose Control ................................................................................................................10

5.3.1 Air Monitoring ...................................................................................................................12

5.3.2 Early Technical and Analytical Capabilities....................................................................... 12

5.3.3 Recent History of Bioassay Results.................................................................................. 13

5.4 Minimum Detectable Activities .................................................................................................. 14

5.5 Development and Impact of Whole-Body Counting (WBC) ....................................................... 15

5.6 Default for Missed Dose............................................................................................................ 18

5.7 Unmonitored Workers ...............................................................................................................18

References ........................................................................................................................................20
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Glossary ............................................................................................................................................24

LIST OF TABLES

Table

5-1 Radionuclides of concern for ANL-W .........................................................................................

Page

8

5-2 ANL-W routine bioassay history summary .................................................................................9

5-3 Internal dose assessment information after mid 1990.............................................................. 10

5-4 Internal dose assessment information before mid 1990........................................................... 10

5-5 Analytical information that could be in claimant dose files........................................................ 11

5-6 Area codes that could be in claimant dose files. ...................................................................... 11

5-7 Surface contamination control and MDAs................................................................................ 12

5-8 ANL-W Recent History of Bioassay Sampling.......................................................................... 14

5-9 MDAs for urine samples by time period ................................................................................... 15

5-10 MDAs for fecal samples by time period.................................................................................... 16

5-11 In vivo MDAs by time period .................................................................................................... 16

5-12 Default table for missed dose .................................................................................................. 19
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 11/09/2004 00-A New technical basis document for the Argonne 

National Laboratory – West – Occupational 

Internal Dose. Initiated by Norman D. Rohrig.

Draft 12/15/2004 00-B Incorporates comments from internal review.

Initiated by Norman D. Rohrig.

Draft 01/11/2005 00-C Incorporates comments from NIOSH review.

Initiated by Norman D. Rohrig.

Draft 01/18/2005 00-D Incorporates additional comments from NIOSH 

review. Initiated by Norman D. Rohrig.

01/20/2005 01/20/2005 00 First approved issue. Initiated by Norman D. 

Rohrig.
Effective Date: 01/20/2005 Revision No. 00 Document No. ORAUT-TKBS-0026-5 Page 4 of 26

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AECL administrative exposure control level

AEDE annual effective dose equivalent

AMAD activity median aerodynamic diameter

ANL-W Argonne National Laboratory-West

ANSI American National Standards Institute

CAM continuous air monitor

CDE committed dose equivalent

CEDE committed effective dose equivalent

CFA Central Facilities Area

CFR Code of Federal Regulations

Ci curie

cm centimeter

COO Chicago Operations Office

CPP Chemical Processing Plant

cpm counts per minute

d day

DAC derived air concentration

DAC-hr derived air concentration-hour 

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

dpm disintegrations per minute

EBR-I Experimental Breeder Reactor No. 1

EBR-II Experimental Breeder Reactor No. 2

ERDA Energy Research and Development Administration

F fast absorption type

FASS Fixed Air Sampling System

FCF fuel cycle facility

hr hour

HFEF Hot Fuel Examination Facility

HP health physicist

H&S Health and Safety

HSD Health and Safety Division

HSL Health Services Laboratory

ICRP International Commission on Radiological Protection

IDO Idaho Operations Office

INEEL Idaho National Engineering and Environmental Laboratory

INL Idaho National Laboratory

L liter

m meter

M moderate absorption type

mCi millicurie
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MDA minimum detectable activity or amount

MDL minimum detectable level

MFP mixed fission products 

mg milligram

min minute

mL milliliter

MPBB maximum permissible body burden

MPC maximum permissible concentration

MPCa MPC for airborne activity

mrem millirem

mrep millirep

NaI(Tl) sodium iodide doped with thallium

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NRTS National Reactor Testing Station

pCi picocurie

RAM radiation area monitor or remote area monitor

RDR Radiation Dosimetry and Records

RESL Radiological Environmental Sciences Laboratory

S slow absorption type

TBD technical basis document

TLV threshold limit value

TRU transuranic

U.S.C. United States Code

WBC whole-body counting

yr year

ZPPR Zero Power Plutonium Reactor (later Zero Power Physics Reactor)

α alpha particle

β beta particle 

γ gamma

σ standard deviation

μCi microcurie

μg microgram

μm micrometer
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5.1 INTRODUCTION AND HISTORICAL OVERVIEW

Technical Basis Documents (TBDs) and Site Profile Documents are general working documents that 

provide guidance concerning the preparation of dose reconstructions at particular sites or categories 

of sites. They will be revised in the event additional relevant information is obtained about the 

affected site(s). These documents may be used to assist the National Institute for Occupational 

Safety and Health (NIOSH) in the completion of the work required for each individual dose 

reconstruction.

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

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

employer facility” or a “Department of Energy (DOE) facility” as defined in the Energy Employees 

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

In 1949, the U.S. Atomic Energy Commission (AEC) established the National Reactor Testing Station 

(NRTS) and started construction of facilities on a 572,000-acre site approximately 50 miles west of 

Idaho Falls in southeastern Idaho. NRTS was later the Idaho National Engineering Laboratory (INEL) 

and is now the Idaho National Engineering and Environmental Laboratory (INEEL). Argonne National 

Laboratory (ANL) near Chicago established a branch at the NRTS known as ANL – West (ANL-W) 

where they built and operated several reactors which were of fundamental importance for the 

development of commercial nuclear power. ANL-W was operated under contract to the Chicago Field 

Office of the AEC/DOE by the University of Chicago, operating contractor from 1951 through January 

2005. ANL-W will be merged with the research side of the INEEL to form the INL and will be operated 

by Battelle Energy Alliance beginning in February 2005. 

Each of the original AEC Laboratories was unique in both mission and location. Since the early days 

of the AEC programs represented the beginnings of the nuclear age, significant technical 

developments were a necessity, not the least of which were developments in radiation safety areas. 

Some of the unique characteristics of radiation safety (and internal dosimetry specifically) at the 

NRTS/ANL-W, which had a marked influence on the conduct of the internal dosimetry programs are 

as follows:

• The original mission of the ANL-W was uranium reactor concept development. The production 

of weapons grade nuclear materials was not a mission.

• The ANL-W beginning was 8-10 years later than ANL-East, Oak Ridge National Laboratory, 

and Hanford. During those developmental years significant technical progress in professional 

skills, instrumentation, analyses, procedures, and techniques were accomplished. 

• The AEC/DOE Chicago field office was responsible for and had oversight of the ANL-W 

program. 

• To provide needed consistency of radiation safety programs at the NRTS, the AEC through 

the Idaho field office established a Health and Safety Laboratory (HSL) to provide technical 

support in the areas of 1) environmental surveillance, 2) external dosimetry (personnel 

dosimeters of all types), 3) portable radiation detection instrumentation inventories, calibration, 

and maintenance, 4) internal in vitro and in vivo bioassay analytical laboratories, 

5) maintenance and documentation of personnel dosimetry records, and 6) research and 

development in these areas of responsibility. The name of this organization changed several 

times to Health Services Laboratory (HSL), Health and Safety Division (HSD), Idaho Center for 

Radiological and Environmental Sciences (ICRES), and now to the Radiological and 

Environmental Sciences Laboratory (RESL). Technical data, dosimetry services, information 
Effective Date: 01/20/2005 Revision No. 00 Document No. ORAUT-TKBS-0026-5 Page 7 of 26

(particularly in the instances of detectable worker intake), and analytical internal dose 

calculations and evaluations were exchanged between the AEC HSL and ANL-W. 

As a consequence, basic assumptions about minimum detectable activities (MDAs) or minimum 

detectable levels (MDLs), missed dose potential, etc. are relatively consistent across the years. As 

early as 1955 or 1956 gamma spectral analysis capabilities allowed the significant bioassay results 

(those which would result in reportable internal dose) to be defined in terms of the specific 

radionuclides. The practice in the case of a higher urine sample result was to attempt radionuclide 

identification through gamma spectral analysis and chemical separation. This document describes

default assumptions for use in cases when the bioassay records for a claimant do not include specific

radionuclide analyses and only record gross beta or gross alpha results.

The majority of this document provides background information to aid the internal dose reconstructor 

through increased general understanding, data interpretation, defaults, etc. Section 5.1.1 provides a 

facility description, and Section 5.1.2 details the radionuclides of concern. Section 5.2 describes the 

INEEL radiological protection program as it evolved over the years, Section 5.3 discusses internal 

dose control, and Sections 5.4 and 5.5 describe MDAs and whole-body counting (WBC), respectively. 

Sections 5.6 and 5.7 suggest approaches for the treatment of missed dose and unmonitored workers.

5.1.1

ANL-W, started in 1951, continues to conduct nuclear power developmental programs. Although 

ANL-W receives internal dosimetry support from the INEEL service laboratories, its radiological safety 

programs operate under authority of the DOE Chicago Operations Office.

Argonne National Laboratory-West

Nine experimental reactors under the technical direction of ANL-W were operated at two locations, 

one on the southwest side of the Site near RWMC and the others at the current location on the 

southeast side of the Site. Early reactor operations included physics critical experiments; power 

production; routine unmoderated operation, uranium-fueled, plutonium-fueled, and breeder reactor 

designs; and self-destruct experiments. 

A series of deliberate safety experiments were conducted by ANL-W in which reactors were allowed 

to go prompt-critical with resultant reactor destruction (Stacy 2000). External and internal doses to 

workers, both expected and accidental, were associated with these activities (RAC 2002). 

5.1.2

ANL-W facilities and activities have been related primarily to experimental reactor design and 

development. The ANL-W Site Description TBD (ORAU 2004) describes these activities in more 

detail. The radionuclides of concern as listed in Table 5-1 from an internal dose standpoint are mixed 

fission product (MFP) (primarily aged), uranium and decay products, and transuranic (TRU) and 

decay products. The uranium has had various enrichments from depleted to at least 68% enriched, 

both used in EBR II. Most internal doses have been identified following an incident rather than as a 

result of routine bioassay measurements. Through the years at ANL-W, plotting urine and fecal 

elimination curves has shown that many nuclides appear to be eliminated slowly, indicating a 

relatively insoluble material. The chemical explanation is that radioactive materials oxidize rapidly, 

forming less soluble compounds. A default assumption of M or S would be appropriate, where ICRP 

68 specifies these for the nuclide. Strontium is an exception; type F should be assumed.

Radioactive Nuclides of Concern and Solubility
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Table 5-1. Radionuclides of concern for ANL-W (Nielsen 1996a). 

Radionuclides Sources/characteristics

H-3 (HTO) EBR-II Reactor Facility & Sodium Components Maintenance Shop

Mn-54, Fe-59

Co-58, Co-60

EBR-II primary source. Levels low and decreasing since decommissioning.

Sr/Y-90 All facilities handling fission products

Cs-134, Cs-137 All facilities handling fission products

U-234, 235, 238 Hot cell, hoods, glove boxes, waste, reactor fuel, research areas

Pu-238, 239 FCF, HFEF, ZPPR, Analytical Laboratory

Am-241 FCF, HFEF, ZPPR, Analytical Laboratory

5.2 RADIOLOGICAL PROTECTION PROGRAM MANAGEMENT AND SUPPORT

The contract with the University of Chicago to operate the ANL-W facilities has not changed in this 54-

yr period. The Chicago Operations Office (COO) provided oversight for ANL-W programs and 

facilities. The ANL-W program used site support services, including internal dosimetry support. 

The personnel dosimetry records have been and are documented and permanently maintained. 

Records about individual facility/contractor field monitoring programs (air-monitoring data, personnel 

contamination records, etc.) were maintained by ANL-W. The field monitoring data were not available 

for use in this report. 

The H&S Laboratories had radionuclide identification capabilities from the early 1960s. Positive 

bioassay results (analyses in which the results exceeded 2σ counting statistics) were normally 

followed by a confirmatory analysis to identify specific radionuclides. In the cases where only gross 

beta or gross alpha bioassay results are available, the results are normally within 2σ. However, if it is 

necessary to evaluate intakes from the gross beta or gross alpha results, the radionuclide defaults 

should be 90Sr/90Y and 239

Pu, respectively. 

5.2.1

Routine bioassay of radiation workers has been conducted since the beginning of operations. 

However, formal documentation of the ANL-W bioassay programs was not found for periods before 

1989. Some of the data sheets on individuals indicate that urine bioassay sampling occurred annually 

from 1959 to 1966. Whole body counts began in 1963, typically annually. Table 5-2 lists the 

reconstructed history of routine bioassay frequency.

Bioassay Programs

5.2.2

Formal or legal internal dose data were maintained by the DOE Health and Safety Division (HSD) in 

individual hard-copy folders until mid 1990, when all technical support service functions, including 

those related to internal dosimetry, were transferred to the INEEL prime contractor. At that time, in 

vitro analytical functions were transferred to an onsite INEEL analytical laboratory. The in vivo

counting laboratory provides support directly through the Radiation Dosimetry and Records (RDR) 

organization, which administers external and internal dosimetry support programs. The subject matter 

expert reviews, validates, and prepares official internal dose assessments. The RDR unit functions 

include documentation and records custodial responsibilities. 

Internal Dose Records
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Table 5-2. ANL-W routine bioassay history summary.

Year Typical frequency Type Groups analyzed/sampled Investigating level Comments Reference

1953-

1960

Annually In vitro urine Radiation workers Frequency is inferred 

from individual data 

sheets.

Individual data 

sheets

Table 5-9 

AEC 1959 

AEC 1961

1961-

1972

Annually In vitro urine

In vivo

Radiation workers Frequency is inferred 

from individual data 

sheets.

AEC 1962a

AEC 1963a

1973-

1981

When intake 

suspected

Fecal Radiation workers

Annual DE >10% quarterly 

standard in ERDA Manual 

Chapter 0524

Reporting Frequency is inferred 

from individual data 

sheets.

AEC 1968 

AEC 1975

Annually In vivo ERDA 1975

1982-

1988

When intake 

suspected

Fecal Radiation workers

CDE >10% quarterly 

standard in DOE 5480.1A

Chapter 11

Reporting DOE 1981

Annual In vivo

Termination In vivo Radiation workers

1989-

1995

Annual In vivo Radiation workers, 

available all employees

ES&H 1989

Radiation Workers 

based on exposure 

potential

In vitro fecal Medium to highest risk

working with plutonium 

contamination

In vitro urine Uranium or tritium workers

Termination In vivo When determined by 

Radiation Fire and Safety 

Engineering

In vitro

New hire In vitro

In vivo

1995-

2004

Appropriate to the 

facility mission, 

potential uptakes.

In vivo Radiation workers that 

enter radiological buffer 

areas or areas of greater 

radiological controls and 

are likely to receive intakes 

resulting in a CEDE of 0.1 

rem or more. Type of 

bioassay based on source 

term. Urine requested 

when pure beta, uranium, 

or TRU was of interest. 

Feces requested primarily 

for uranium and TRU 

source terms.

In accordance with DOE 

5480.11 and 10 CFR 835. 

Reporting

Workers that could 

receive 0.1 rem CEDE.

Declared pregnant 

workers when 

embryo/fetus could 

receive 0.05 rem DE. 

Follow-up for any 

suspected intake of 

radionuclides and to 

more accurately 

identify and 

characterize the 

amount of intake and 

excretion pattern.

Nielson 1996a

In vitro DOE 1988

Urine

Fecal

When work place 

monitoring indicates 

significant potential 

for intakes.

Internal doses resulting 

from all confirmed intakes

are to be evaluated.

Investigating

1995-

2004 

New hire To determine internal 

conditions from previous 

uptakes or to establish 

baseline for those 

continuing to work as 

radiation workers.

Termination In vivo All employees

AEDE = annual effective dose equivalent; ERDA = Energy Research and Development Administration.

5.2.3

The changing standards and regulations influenced the level of internal dose evaluation and 

documentation, but did not change the fact that all (negative as well as positive) bioassay data were 

recorded in the individual dosimetry files.

Data Codes and Investigation Levels 

The information used in internal dose assessments and analytical data sheets has varied through the 

years. Table 5-3 describes coded information that could appear in records from after mid 1990. 

Table 5-4 describes internal dose information that could appear in records from before mid 1990. 

Table 5-5 contains analytical nomenclature. Table 5-6 contains INEEL codes for various site areas. 
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Table 5-3. Internal dose assessment information after mid 1990. 

Coded information Description

Name & SS No. Exposed employee by name and social security number.

Asmt. Nos. This assessment number is the calendar year, and a consecutive numbered 

assessment for that employee during that specific year.

Intake Date Month/Day/Year of employee intake.

Radionuclide Class & 

Amt.

Specific radionuclide followed immediately by ICRP Publication 30 solubility class 

symbol D, W, or Y (ICRP 1979). Amount in microcuries or becquerels.

CEDE rem Calculated CEDE in rem.

Organ (Max.) Organ that received the maximum dose from the specified intake.

Organ CDE rem CDE calculated for the listed organ in rem.

Employer and Exp. 

Location

Abbreviation of DOE site contractor and the plant site of exposure (can include the 

building number).

Year – Total CEDE CEDE exposures are summed for the year of intake for each employee.

Year – TL Organ CDE Organ CDE total (TL) exposures are summed for the year of intake for each 

employee.

Table 5-4 Internal dose assessment information before mid 1990. 

Dose information Description

Name, Soc. Sec. No. Employee name, social security number, and (Contractor Abbreviation/Plant or 

Facility).

Nuclide Radionuclide symbol followed by ICRP solubility class (D, W, or Y) (ICRP 1979).

Intake Period Month and Year for single exposure or period of time by month and year in which 

exposure occurred.

CEDE rem Calculated CEDE in rem.

AEDE rem Calculated AEDE in rem.

Year Year for which the AEDE was calculated.

Each individual analytical result was documented and placed in individual exposure files, regardless 

of the formal reporting requirements. The investigation levels (the levels at which positive bioassay 

results triggered follow-up sampling to verify that detectable activity had been taken into the body) 

have also changed little from the early years to the present. Later procedures (DOE 1988) set 

specific limits on those positive bioassay results that could result in 100 mrem annual effective dose 

equivalent or above as the point at which followup and reporting was required. With the DOE 

Radiological Control Manual (DOE 1994), this changed to 100 mrem CEDE. In addition, a calculated 

dose of 10 mrem or above would be recorded as an internal dose. These procedural limits did not 

materially affect the bioassay sampling frequency and the recording of even nondetectable 

radioactivity in bioassay samples, although the request for and number of followup samples and 

analyses could have been different as a function of the formal regulations in effect.

5.3 INTERNAL DOSE CONTROL 

The radiological protection program was designed to detect barrier or ventilation failure in a timely 

manner. The program consisted of continuous and retrospective air and effluent monitoring combined 

with personnel and surface contamination monitoring. Detection of barrier failure provided the 

information for making decisions on evacuating personnel, increasing personnel protection equipment 

(e.g., respirators), and requesting bioassay analyses to identify possible internal intake. 
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Table 5-5. Analytical information that could be in claimant dose files. 

Analytical information Description

Sample No. Sample log number.

Date and Time Generally clear interpretation.

Sample Description Name of the employee, numerical sample number frequently included, additional 

special analyses performed (e.g., Sr-90, Y Separation, etc.).

* Anal. For Generally gross beta and/or gross gamma. Sample aliquots evaporated for gross 

beta or counted directly in a deep well NaI scintillation counter, Specific isotopic 

analysis, based upon chemical separation or gamma spectrum also listed in this 

column.

Quantity Used Size of the sample aliquot – generally in mL.

U+ or K+ Trans. Note to indicate analytical correction for natural potassium and uranium.

Count Time Counting either used preset time or preset counts. Time in minutes recorded in 

either case.

Total Count Total number of counts recorded.

Gross Count, cpm Cpm determined by dividing total counts by time of count.

Bkgd., cpm Background cpm recorded.

Net count, cpm Gross cpm minus background cpm.

K-40 corr., cpm Additional background from K-40 identified. K-40 is not a facility occupational 

exposure product; ignore from an internal dose reconstruction

Foreign Activity, cpm 

and dpm

Net counts corrected for K-40 and then converted to dpm based upon counter 

calibration. Uncertainty also included, which is recorded as 1σ based upon 

counting statistics.

dpm per a volume The activity is for the sample volume listed.

Result in μg/L These results are for uranium whether stated or not.

Table 5-6. INEEL and ANL-W area codes that could be in claimant dose files. 

Area code Description Area code Description

1 AEC Headquarters Bldg 20, 261, 264 TREAT

2 EBR – I 21 LX

3, 034, 035 CFA 22 GCRE

4, 042,045 MTR, TRA 23 OX

5, 053, 055 ICPP 24 ARHG

6 NRF 25 No information available

7 TAN (GE) 26, 263,265 EBR -II

8 Services 27 ML-1

9 NX (X is construction) at NRF 28 On-Site Site Survey

10 AX at TAN 29 Off-Site Site Survey

11, 113 CX at CPP 30 ANP Program at SL-1

12 EX at EBR 31 STPF

13, 133,135 SPERT, PBF 65 ECF

14 MORE 66 Non-Security

15 SX at SPERT 67 Division of Compliance

16 SL-1 68 STEP

17, 333 MX at MTR 69 LPTF (Phillips & AEC)

18, 814,815 WP, RWMC 71 CADRE

19, 772, 775 TAN (Phillips & AEC) 774,776 SMC

Bold is for ANL-W facilities.

The contamination control limits for the detection and control of released activity were dependent on 

instrumentation capabilities. As a result of increased emphasis on exposures that were as low as 

reasonably achievable, some reduction in acceptable release levels was implemented. The 

contamination control limits for alpha on plant surfaces and particularly on personnel were set close to 
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the MDA, such that any detectable contamination was a signal for preventative action and follow-up 

evaluations. Beta/gamma MDAs typically were a factor of 5 below the limits. Table 5-7 is a summary 

of control limits primarily from the CPP [Chemical Processing Plant] Health Physics Manual (ACC 

1952) and current operating procedures.

Table 5-7. Surface contamination control and MDAs. 

Period Surface location

Detection 

technique Control levels MDA - typical

1952-

1960s

Plant/equipment Smears 500 dpm β & 20 dpm α per 

100 cm2 150 dpm β & 10 dpm α per 100 cm2

Personal clothing Portable survey 

instruments

1500 dpm β & 500 dpm α

per 100 cm2

1,000 dpm β & 500 dpm α

per 100 cm2

Personal skin Portable survey 

instruments

Any detectable reported, 

e.g. 1,000 dpm β & 500 

dpm α per 100 cm2 

1,000 dpm β & 500 dpm α

per 100 cm2

Shipments Smears/portable 

survey 

instrument

500 dpm β & 20 dpm α per 

100 cm2

 – smears

0.1 mrep/hr β & 500 dpm α

per 100 cm2

150 dpm β & 10 dpm α per 

100 cm2 smears. 0.01 

mrep/hr β & 500 dpm α per 

100 cm2

1970spresentPlant/equipment 

surfaces

Smears 300 dpm β & 20 dpm α per 

100 cm2 30 dpm β & 10 dpm α per 100 cm2

Personnel Portable survey 

instruments

Any detectable reported, 

e.g. 300 dpm β & 200 dpm

α per 80-100 cm2 

300 dpm β & 200 dpm α per 

80-100 cm2

5.3.1

The monitoring of radioactivity in the air in occupied areas was a basic element of the internal 

exposure prevention program. Beta/gamma continuous air monitors (CAMs) were used for 

beta/gamma emitters with maximum permissible concentrations/derived air concentrations 

(MPCs/DACs) above 1 × 10-9 μCi/cm3

. TRU materials and uranium are/were nearly always well 

tagged with beta/gamma activity that allowed beta/gamma-detecting CAMs to be used to warn of 

possible alpha contamination and possible internal exposures. 

Air Monitoring

In general, workers were asked to submit to bioassay whenever they were in an area where a CAM 

alarm sounded. In addition, the fixed location and retrospective air-sampling system would signal the 

need for bioassay, if elevated air sample results were detected.

At ANL-W fixed air sampling heads were used in the Fixed Air Sampling System (FASS) beginning in 

1976 (Courtney 1980). Concentrations have averaged between 1 and 5 x 10-15 μCi cm-3 for alpha 

emitters and between 0.4 and 4.0 x 10-13 μCi cm-3 for beta emitters in the first three years of operation. 

This information could be used to assign an intake for an unmonitored worker.

5.3.2

DOE HSL technical reports and annual reports, coupled with facility memoranda and reports, 

documented the analytical detection capability in the 1950s and 1960s. Radionuclide identification by 

their energy spectra was available and used for urine and other bioassay samples. Specific 

separations (e.g., strontium, iodine) were available to quantify the radioactive components of a variety 

of samples of interest.

Early Technical and Analytical Capabilities
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In the early days a gross beta urine bioassay measurement was made on an evaporated aliquot or a

gamma count was made directly on a liquid sample, or both. Any detectable activity triggered a 

specific chemical separation analysis (generally strontium). Early analyses for plutonium generally 

were gross alpha counts on a plutonium separation; later, alpha spectroscopy was used to count and 

better characterize the results.

In 1958, the IDO H&S Division acquired a 256-channel gamma spectrometer with a 3- by 3-in. sodium

iodide thallium-doped [NaI(Tl)] detector counting system for analyses of gamma-emitting 

radionuclides. In 1960, the HSD obtained a 3- by 3-in. well counter for gamma analysis, replacing the 

previously used gross beta counting as the routine analytical procedure for urine samples. AEC 

(1961, p. 59) states, “Approximately 1.5 × 10-6 μCi/mL of MFPs can be detected in 75 mL of urine in a 

5-minute count which is about the same as was obtained with the gross beta procedure in a 20-

minute count”

AEC (1961) outlined a basic philosophy in relation to gamma counting of bioassay samples. Gamma 

counting would be effective in all situations except for exposure to pure strontium isotopes. To guard 

against this unlikely possibility, the procedure of performing a strontium analysis for individual workers 

at risk (radiation workers) every 2 yr and at termination was established. Because of the improbability 

of finding detectable activity, all activities were to be precipitated by oxalic acid in a weak acid 

solution, gross beta counted, and the strontium specific analysis not completed unless a detectable 

count was obtained on the precipitate. A 100-mL sample of urine permitted the detection of 

approximately 8 × 10-8 μCi/mL of 90Sr.

The special and routine bioassay sample analyses were performed and documented by the DOE 

analytical laboratory. The individual analytical data were recorded. At ANL-W, there is a personnel 

dose file on each individual which contains results of all samplings and notations on all contamination 

incidents. In all cases, copies of the bioassay results should be in the individual’s dose file in the 

NIOSH Office of Compensation and Support Claims Tracking System (NOCTS) database.

5.3.3

Table 5-8 provides a recent history of bioassay sampling based on annual skin contamination and 

exposure reports (Carr 1990; Holson 1991; Kirchner 1992; Marshall 2002; Nielsen 1992a; Nielsen 

1992b; Nielsen 1993a; Nielsen 1993b; Nielsen 1994a; Nielsen 1994b; Nielsen 1995a; Nielsen 1995b; 

Nielsen 1996b; Nielsen 1996c; Nielsen 1997a; Nielsen 1997b; Nielsen 1998a; Nielsen 1998b; Nielsen 

1999a; Nielsen 1999b; Nielsen 2000a; Nielsen 2000b; Nielsen 2003; Nielsen 2004; Vroman 2001). 

Blank spaces mean that the information was not available. The skin contaminations tend to be a 

precursor for positive bioassay and the number has been countable and tracked since at least 1984. 

The activities associated have not been extremely high. In 1993 there were a large number of skin 

and clothing contaminations associated with the Fuel Cycle Facility (FCF) which was being 

refurbished. Almost everybody received a whole body count until the late 1990s. In 1998, a FCF seal 

tube repair led to 11 positive WB counts on 5 people. The highest calculated CEDE was 1 mrem. 

The other positive count was from a medical test about 3 years previous. The tritium assay was about 

monthly so many fewer people were involved and the doses were generally trivial because of the high 

monitoring sensitivity. The number of tritium assays was cut back because of the history of not seeing 

anything significant in dose from tritium. The Pu fecal assay led to mostly false positive results where 

a subsequent fecal assay failed to show any activity. In 1993 a person received a hand wound and 

was chelated resulting in two positive urine counts but no positive fecal count (Burke 1993). A 100 

mrem CEDE was assigned. This recent history suggests rather modest internal exposures occurred 

at ANL-W.

Recent History of Bioassay Results
Effective Date: 01/20/2005 Revision No. 00 Document No. ORAUT-TKBS-0026-5 Page 14 of 26

Table 5-8. ANL-W Recent History of Bioassay Sampling. 

Year

# Skin 

Contam 

Events

#Clothng 

Contam 

Events

Maximum 

Beta 

Contam

# Whole 

Body 

Counts

Positive 

WBC

# Fecal 

Pu 

Assay Pos in vitro

# Tritium 

Assay

Pos 

Tritium

Max T 

Dose

1984 20 cpm till 

1992

mrem

1985 17

1986 7 dpm after

1987 5 1992

1988 16

1989 18 8000

1990 10 542 0 44 0

1991 10 12000 648 0 60 0 287 29 14

1992 15 7 8000 609 0 69 0 40 8 <1

1993 23 58 500000 1082 0 33 2 Urine 48 6 <1

1994 7 21 139000 837 0 70 1 FPa 46 14 <1

1995 3 10 784000 758 0 106 1 FP 106 10 <1

1996 2 11 186000 273 0 75 1 FP 139 13 <1

1997 5 10 135000 329 1 76 2 FP 1 Pa 138 8 <1

1998 5 21 309 12 56 2 P 152 16 <1

1999 7 5 311 1 105 2 FP 365 84 <1

2000 2 4 318 1 136 0 697 50 <1

2001 1 7 254 1 143 1 P 552 73 <1

2002 4 8 212 1 143 3 P 40 7 <1

2003 2 8 222 0 127 1 FP 9 1 <1 a

FP – false positive P - Positive

5.4 MINIMUM DETECTABLE ACTIVITIES

In compliance with the November 1998 Code of Federal Regulations requirement (10 CFR 835) for 

DOE Laboratory Accreditation Program (DOELAP), and based on American National Standards 

Institute (ANSI) N 13.30, Performance Criteria for Radiobioassay (ANSI 1996), both the in vitro and 

in vivo radiobioassay laboratories at INEEL received DOELAP accreditation in February 1998. In 

accordance with this accreditation, MDAs and decision levels at the 95% confidence level are 

performed. Tables 5-9 and 5-10 list current MDAs for urine and fecal sample analysis, respectively, 

along with values gleaned from historical documents. The earliest reference (Ebersole and Flygare 

1957) defines a detection limit as “twice the standard deviation of its determination”. Thus the 

detection limits in early references have been multiplied by 2.33 to determine an MDA similar to the 

ANSI defined values. A large majority of the urine samples taken were single voidings; 24-hr samples 

were used for special sampling purposes (i.e., followup samples, primarily to extend the sensitivity). 

The MDA for urine sample is given in both pCi/ml and in pCi/sample if the sample aliquot size is 

stated, and if the size is not stated only the reference value is provided. Most bioassay data has units 

of activity/sample with the units of the activity changing from dpm to μCi to Bq and back again. The 

MDAs listed are those for the primary samples. The recommended time periods for the MDA values 

have been included in the tables. Table 5-11 lists the current in vivo MDAs along with values gleaned 

from historical documents and the recommended periods for use of the MDA values.
Effective Date: 01/20/2005 Revision No. 00 Document No. ORAUT-TKBS-0026-5 Page 15 of 26

Table 5-9. MDAs for urine samples by time period. 

Radiation/

radionuclide Time Period 

Urine

(pCi/mL)

Urine

(pCi/sample)

Nominal Sample 

Volume(mL) Reference

Gross β 1951-1953 8.4 Data Sheet

1954-1959 9.3 46 5 Ebersole & Flygare 1957a

1960-70 3.5 AEC 1961a

1971-present 1 5 5 AEC 1972a

Gross γ 1960-1964 3.5 260 75 AEC 1961a

1965-1971 2.3 Data Sheet

1972-present 4.6 345 75 AEC 1972a

H-3 1972-1994 35 105 3 AEC 1972a , AEC 1974

1995-present 1.4 Andersen, Perry, and Ruhter 

1995, Rielly 2001

Co-60 1963-present 0.01 Rich 1990

Sr-90 1953-1959 9.2 Ebersole 1957

1960-1970 0.2 20 100 AEC 1961

1971-1989 0.023 1.7 75 AEC 1972a, AEC 1974a

1990-1994 0.01 Rich 1990

1995-present 1.9 Andersen, Perry, and Ruhter 1995

I-131 1963-present 0.01 Rich 1990

Cs-134 1963-present 0.01 Rich 1990

Cs-137 1963-present 0.01 Rich 1990

Th-230 1974-present 4.6 E-5 0.046 1000 AEC 1974 a

Np-237 1974-present 4.6 E-5 0.046 1000 AEC 1974 a

U (FP) 1954-1985 14 μg/L Assumed value

U (KPA) 1985-present 0.2 μg/L Rich 1990

U-233/234 1970-1994 1.E-4 Rich 1990

1995-present 0.041 Andersen, Perry, and Ruhter 1995

U-235 1970-1994 1.E-4 Rich 1990

1995-present 0.038 Andersen, Perry, and Ruhter 1995

U-238 1970-1994 1.E-4 Rich 1990

1995-present 0.030 Andersen, Perry, and Ruhter 1995

Pu-238 1974-1989 2.3 E-5 0.023 1000 AEC 1974 a

1990-1994 6 E-5 Rich 1990

1995-present 0.022 Andersen, Perry, and Ruhter 1995

Pu-239/240 1964-1970 4.E-4 AEC 1964a

1971-1973 5.E-4 0.5 1000 AEC 1972a

1974-1989 2 E-5 0.02 1000 AEC 1974 a

1990-1994 6.E-5 Rich 1990

1995-present 0.027 Andersen, Perry, and Ruhter 1995

Am-241 1974-1989 7 E-5 0.07 1000 AEC 1974 a

1990-1994 2. E-4 Rich 1990

1995-present 0.023 Andersen, Perry, and Ruhter 1995

Cm-244 1974-present 1.4 E-5 0.014 1000 AEC 1974 a

Cf-252 1974-present 1.4 E-5 0.014 1000 AEC 1974 a

a. MDA calculated from inferred 2σ uncertainty.

5.5 DEVELOPMENT AND IMPACT OF WHOLE-BODY COUNTING (WBC)

WBC was introduced at the INEEL in 1961. As early as 1961 one of the fundamental conclusions 

from the experience with in vivo and in vitro internal dosimetry analytical techniques was the fact that 

a large proportion of the internal exposures was to insoluble materials. Radioactive nuclides (e.g. 

125Sb, 110mAg, 65Zn, and 95Zr/Nb) were detected by an in vivo count and were not detected in the urine. 

Concurrent analyses of fecal and urine excreta demonstrated the main elimination route to be by the 

feces, with so little voided in the urine as to be undetectable even in a 24-hr specimen (AEC 1962a; 
Effective Date: 01/20/2005 Revision No. 00 Document No. ORAUT-TKBS-0026-5 Page 16 of 26

Sill, Anderson, and Percival 1964). WBC was demonstrated to detect activity as low as 0.01 μCi in a 

10-min count (AEC 1962a). This detection level was several orders of magnitude more sensitive than 

the maximum permissible body burdens (MPBB) for most beta/gamma fission and activation products. 

As a consequence, the in vivo counting program was used to count 1) all terminating employees, 2) 

employees suspected of having a possible internal intake, and 3) any interested employee on an 

annual basis. Only those activities greater than 0.1 μCi were further quantified. This level was 

determined to be less than one-tenth of the MPBB for most of the gamma-emitting isotopes.

Table 5-10. MDAs for fecal samples by time period. 

Radiation/

radionuclide Time Period

Fecala

(pCi/sample) Reference

Co-60 1963-present 10 Rich 1990

Sr-90 1963-1994 10 Rich 1990

1995-present 1.9 Andersen, Perry, and Ruhter 1995, Rielly 2001

Cs-134 1963-present 10 Rich 1990

Cs-137 1963-1999 10 Rich 1990

2000-present 0.3 Bechtel BWXT 2000

Th-230 1974-present 0.03 AEC 1974

Np-237 1974-present 0.03 AEC 1974

U-233/234 1970-2002 0.041 Andersen, Perry, and Ruhter 1995, Rielly 2001

2003-present 0.05 Bhatt 2003

U-235 1970-2003 0.038 Andersen, Perry, and Ruhter 1995, Rielly 2001

2003-present 0.09 Bhatt 2003

U-238 1970-1994 0.5 Rich 1990

1995-2002 0.03 Andersen, Perry, and Ruhter 1995, Bechtel BWXT 2000, Rielly 

2001

2003-present 0.09 Bhatt 2003

Pu-238 1974-1994 0.03 AEC 1974

1995-2002 0.022 Andersen, Perry, and Ruhter 1995, Rielly 2001

2003-present 0.02 Bhatt 2003

Pu-239/240 1964-1973 0.4 AEC 1964b

1974-1994 0.02 AEC 1974

1995-present 0.03 Andersen, Perry, and Ruhter 1995, Bechtel BWXT 2000, Rielly 

2001, Bhatt 2002

Am-241 1974-1994 0.07 AEC 1974

1995-2001 0.023 Andersen, Perry, and Ruhter 1995, Rielly 2001

2002-present 0.04 Bhatt 2002

Cm-244 1974-present 0.02 AEC 1974

Cf-252 1974-present 0.02 AEC 1974

a. When sample size is not identified in individual’s records, assume the activity is that excreted per day.

b. MDA calculated from inferred 2σ uncertainty.

Table 5-11. In vivo MDAs by time period. 

Radiation/

radionuclide Time Period

In vivo result

(nCi)

Count time 

(min) Reference

Cr-51 1962-2000 12 10 Percival and Anderson 1962a

2001-present 32 5 Rielly 2001

Mn-54 1962-2000 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, Ruhter 1995

2001-present 2.6 5 Rielly 2001

2001-present 1.3 10 Rielly 2001

Fe-59 1962-2001 4.5 5 Rielly 2001

2001-present 1.5 10 Rielly 2001

Co-58 1962-2000 12 10 Percival and Anderson 1962a

2001-present 2.5 5 Rielly 2001

2001-present 1.1 10 Rielly 2001

Co-60 1962-1970 12 10 Percival and Anderson 1962a

1971-1988 5 10 AEC 1972a; AEC 1974
Effective Date: 01/20/2005 Revision No. 00 Document No. ORAUT-TKBS-0026-5 Page 17 of 26

Radiation/

radionuclide Time Period

In vivo result

(nCi)

Count time 

(min) Reference

1989 7 10 Martin 1989

1990-1992 2 (lung) Rich 1990

1993-2000 7 10 Grothaus 1993; Andersen, Perry and Ruhter 1995

2001-present 2.5 5 Rielly 2001

2001-present 1.1 10 Rielly 2001

Zn-65 1962-1988 12 10 Percival 1962a

1989-2000 10 10 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter

1995

2001-present 4.9 5 Rielly 2001

2001-present 2 10 Rielly 2001

Sr/Y-90 1968-1977 70 (skull) 10 AEC 1969a; AEC 1972a; AEC 1974

1978-present 34 (skull) 10 Martin 1989; Grothaus 1993

Zr/Nb-95 1962-1988 12 10 Percival and Anderson 1962a

1989-2000 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter

1995

2001-present 2.6 5 Rielly 2001

Ru-106 2001-present 27 5 Rielly 2001

2001-present 7.6 10 Rielly 2001

Ag-110m 1962-present 12 10 Percival and Anderson 1962a

Sb-125 1962-present 14 10 Martin 1989; Grothaus 1993

I-131 1962-1989 12 10 Percival and Anderson 1962

1990-1992 2 (thyroid) 10 Rich 1990

1993-2000 0.3 (thyroid) 10 Grothaus 1993

2001-present 3.8 5 Rielly 2001

2001-present 0.13 (thyroid) Rielly 2001

Cs-134 1989-2000 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter

1995

1990-present 2(lung) Rich 1990

2001-present 3 5 Rielly 2001

2001-present 0.96 10 Rielly 2001

Cs-137 1962-1970 12 10 Percival and Anderson 1962a

1971-1998 5 10 AEC 1972a; AEC 1974; Martin 1989; Grothaus 1993;

Andersen, Perry, and Ruhter 1995

1999-2000 2. (lung) 10 Rich 1990

2001-present 3.1 5 Rielly 2001

2001-present 1.9 10 Rielly 2001

Ba/La-140 1962-present 12 5 Rielly 2001

Ce-141 1962-present 9.9 5 Rielly 2001

2001-present 3.2 10 Rielly 2001

2001-present 0.11 (lung) 60 Reilly 2001

Ce-144 1962-2000 50 10 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 

1995

2001-present 44 5 Rielly 2001

2001-present 15 10 Rielly 2001

2001-present 0.44 (lung) 60 Rielly 2001

Eu-152 1962-present 4 10 Rielly 2001

2001-present 0.18 (lung) 60 Rielly 2001

Eu-154 1962-present 2 10 Rielly 2001

Eu-155 1962-present 1 10 Rielly 2001

Ga-153 1962-present 6.5 10 Rielly 2001

2001-present 0.096 (lung) 60 Rielly 2001

Hf-181 1962-present 5 10 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter

1995

Ta-182 1962-present 12 10 Percival and Anderson 1962a

Hg-203 1962-present 12 10 Percival and Anderson 1962a

Th-230 1974-present 1,000 AEC 1974

Th-234 2001-present 1.4 (lung) 60 Rielly 2001

Np-237 1974-present AEC 1974

U-235 1993 0.2 (wound) 20 Grothaus 1993
Effective Date: 01/20/2005 Revision No. 00 Document No. ORAUT-TKBS-0026-5 Page 18 of 26

Radiation/

radionuclide Time Period

In vivo result

(nCi)

Count time 

(min) Reference

1962-present 0.2 (lung) Rich 1990

2001 0.11 (lung) 60 Rielly 2001

U-dep/nat 1989 3 (lung) 60 Martin 1989; Grothaus 1993

Pu-238 1989-1998 26 (lung) 60 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 

1995

1993 1 (wound) 20 Grothaus 1993

1999-2000 30 (lung) Rich 1990

2001 54 (lung) 60 Rielly 2001

Pu-239/240 1971-1993 30 100 AEC 1972a

1974-1988 74 (lung) 100 AEC 1974

1989-1988 80 (lung) 60 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 

1995

1993 2 (wound) 20 Grothaus 1993

1990-2000 30 (lung) Rich 1990

2001-present 140 (lung) 60 Rielly 2001

Am-241 1989-1999 0.6 (lung) 60 Martin 1989; Grothaus 1993; Andersen, Perry, and Ruhter 

1995

1993 0.1 (wound) 20 Grothaus 1993

1990-2000 0.2 (lung) Rich 1990

2001-present 0.14 (lung) 60 Rielly 2001

a. MDA calculated from inferred 2-σ uncertainty.

5.6 DEFAULT FOR MISSED DOSE

Based on the ANL-W operational environment characteristics, a number of missed dose default 

assumptions have been derived. The breeder reactors have a fast neutron spectrum and stainless 

steel clad fuel. Rather than include all of the radionuclides in the default summary table for missed 

dose, only 90Sr, 137Cs, 144Ce, and 239

Pu are included for stainless steel fuels. Cesium-137 was 

selected because it is most commonly reported in the in vivo results rather than for its dose 

contribution. The potential missed inhalation dose from the other radionuclides is accounted for by 

weighting the dose from these selected radionuclides by the weighting factors. This gives an 

equivalent to 100% of the dose from the radionuclide distributions for the three types of fuels. Table 5-

12 is a summary of these recommended defaults. 

For most of the history of the ANL-W facilities, personnel dosimeters were issued to all workers who 

entered the security access control points at each facility, regardless of work assignments. For 

example, administrative and clerical personnel were required to wear these radiation-monitoring 

dosimeters even though they were not exposed to elevated backgrounds or internal dose potential. 

Moreover, whole body counting was made available to any worker irregardless of their likely exposure 

to radiation. If exposure was likely, the worker was requested to have a whole body count (ES&H 89). 

5.7 UNMONITORED WORKERS

If no detectable external dose or whole body count information is recorded and no in vitro samples

were recorded, the person should be considered an unmonitored worker for internal dose purposes 

and only the environmental dose should be included.

As noted above, ANL-W personnel dosimeters were issued to all workers. Many of these workers, 

due to the nature of their work environment, would not have inhaled or ingested radioactivity and 

therefore would not have been subjected to routine bioassay.

Most radioactivity encountered was well tagged with β/γ activity, which produced measurable direct 

radiation doses on a personnel dosimeter. Therefore, the probability that a worker with no external 
Effective Date: 01/20/2005 Revision No. 00 Document No. ORAUT-TKBS-0026-5 Page 19 of 26

dose received a significant unmonitored internal intake of radioactive material is very low. Individuals 

who were not issued a personal dosimeter and have no record of internal dose monitoring should be 

assigned only the environmental dose for the facility.

At ANL-W each construction job was evaluated to determine if radiation exposure or internal dose 

could be received. When construction work was done in an area with potential radiation exposure, 

including internal dose exposure, construction workers were monitored in the same fashion as a 

radiation worker. Construction workers who were issued personnel dosimeters should be treated the 

same as facility employees who were issued personnel dosimeters. Construction workers who were 

not issued a personnel dosimeter should be assigned the environmental dose for the facility.

Table 5-12. Default table for missed dose.

Period Based on Recommendation Basis

Start up date 

through 1960

Urine

gross β

Calculate chronic Sr-90 intake that 

results in a urine activity of 0.4 × gross β

Typical β activity is 0.33 Sr-90, 0.33 Y-90 & 0.33 

Cs-137. Use of 0.4 is claimant favorable

Cs-137 intake = Sr-90 intake Half-lives and fission yields of Cs-137 & Sr-90 are 

approximately equal.

Pu-239 intake = 0.004 × Sr-90 intake Pu:Sr-90 ratio of 0.003 weighted by a factor of 1.2. 

Ce-144 intake = 2.4 × Sr-90 intake Ce-144:Sr-90 ratio of 2 weighted by a factor of 1.2.

1961-1980 In vivo

Cs-137

Calculate chronic Cs-137 intake that 

results in the in vivo measurement

Sr-90 intake = Cs-137 intake when no in 

vitro measurement

Half-lives and fission yields are approximately 

equal.

Pu-239 intake = 0.004 × Cs-137 intake 

when no in vitro measurement

Pu:Cs-137 ratio of 0.003 in stainless steel fuel 

weighted by a factor of 1.2. 

Ce-144 intake = 2.4 × Cs-137 assigned 

intake when no measurement

Ce-144:Cs-137 ratio of 2 in stainless steel fuel 

weighted by a factor of 1.2.

1981 - present Bioassay Use bioassay results.

 
Effective Date: 01/20/2005 Revision No. 00 Document No. ORAUT-TKBS-0026-5 Page 20 of 26

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GLOSSARY

absorption

As used in this internal dosimetry section, absorption refers to the material being 

transported to fluids and other organs as well as radiation energy being imparted. 

activation

The process of inducing radioactivity by irradiation.

Atomic Energy Commission

An agency established by the U.S. Government for oversight of nuclear weapons and 

power production; a predecessor to the U.S. Department of Energy.

becquerel

A unit of radioactivity equal to one-disintegration per second.

beta radiation

Radiation consisting of electrons or positrons emitted at high velocity from the nuclei of 

certain radioactive elements. Most direct fission products emit beta radiation. 

breeder reactor

A nuclear reactor concept in which the operation produces a net increase in fissionable 

material.

cladding

The outer layer of material encasing a reactor fuel element (e.g., aluminum or 

zirconium). Cladding promotes the transfer of heat from the fuel to the coolant and 

contains fission products and activation products that result from the fissioning of the 

fuel.

contamination, radioactive

Radioactive material where it does not belong. 

core

That part of the reactor consisting of the fuel and some of the control elements for 

reactor operation.

curie

A unit of radioactivity equal to 3.7x1010 disintegrations per second.

decontaminate

Removing a contaminant, such as a radioactive material, from an undesired location.

dosimeter

A device used to measure accumulated radiation exposure. 

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, 

etc., from external or internal sources of radiation. 

fission

A nuclear transformation characterized by the splitting of a nucleus into at least two 

other nuclei and the release of a relatively large amount of energy.
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fission product

Radionuclides resulting from fission.

fuel reprocessing

A chemical process, usually involving several steps, that recovers 235U and other 

fissionable products from spent fuel.

gamma rays

Short wave length electromagnetic radiation (photons) originating in atomic nuclei and 

accompanying many nuclear reactions (e.g., fission, radioactive decay, and neutron 

capture) in an energy range of 10,000 electron volts to 9 million electron volts. 

half-life

The time it takes for one-half of any given number of unstable atoms to decay 

(disintegrate).

ionizing radiation

Electromagnetic or particulate radiation capable of producing charged particles 

through interactions with matter.

in vitro

In glass. Outside the living body and in an artificial environment. Internal bioassay 

sampling, such as fecal samples or urine samples.

in vivo 

In the living. In the living body of a plant or animal. Bioassay sampling by whole-body 

counting 

isotope

Nuclides having the same number of protons in their nuclei (same atomic number), but 

having a differing number of neutrons (different mass number).

millirem

A unit of radiation dose equivalent (or equivalent dose) equal to one-thousandth of a 

rem (see rem).

microcurie

A measure of radioactivity equal to one-millionth of a curie. 

mixed waste

Waste that contains hazardous and radioactive materials. 

natural uranium

Uranium that has not been through an enrichment process.

neutron

A basic particle in a nuclear reaction, electrically neutral, with nearly the same mass as 

a 1

H atom.

nuclear waste

A general term used for the byproduct unusable material resulting from nuclear 

reactions, including high-level, intermediate, low-level, mixed and TRU waste.
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nucleus

That part of an atom consisting of the total positive electrical charge and most of the 

mass. 

quality factor, Q

A modifying factor used to derive dose equivalent from absorbed dose.

radiation

Energy transferred through air or some other media in the form of particles or waves 

(see ionizing radiation).

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma or 

X-rays, or neutrons from unstable atoms.

radionuclide

A radioactive species of an atom characterized by the constitution of its nucleus 

specified by atomic number (the number of protons), and the mass number (equal to 

the number of protons plus neutrons). 

rem

A unit of dose equivalent, equal to the product of the absorbed dose and the quality 

factor. The word derives from roentgen equivalent in man. 

shielding

Any material or obstruction that absorbs (or attenuates) radiation to protect personnel 

or materials.

spent nuclear fuel

Reactor fuel containing fission and activation products that can no longer economically 

sustain a chain reaction.

Type

Refers to the rate of absorption from lung to blood of inhaled radioactive materials and 

includes types F (fast), M (moderate), and S (slow).

transuranic (TRU) waste

Contaminated waste materials with nuclides having an atomic number greater than 92, 

a half-life over 20 yr and concentration greater than or equal to 100 nCi/g. 

X-ray

Ionizing electromagnetic radiation of extranuclear (outside the nucleus) origin.

zirconium

A metallic element highly resistant to corrosion and often used to make cladding for 

nuclear fuel. It is sometimes alloyed in small amounts in the fuel itself.