ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document for the Idaho National Engineering and 

Environmental Laboratory – Occupational Internal Dose

Document Number:

ORAUT-TKBS-0007-5 

Effective Date: 08/11/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 49

Subject Experts: Bryce L. Rich and Doug Wenzel

Approval: Signature on File Date: 08/11/2004

Norman Rohrig, TBD Team Leader

Approval: Signature on File Date: 08/11/2004

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 08/11/2004

Richard E. Toohey, Project Director

Approval: Signature on File Date: 08/11/2004

James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section 

List of Tables .......................................................................................................................................

Page

2

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

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

5.1 Introduction and Historical Overview...........................................................................................7

5.1.1 Test Reactors ......................................................................................................................8

5.1.2 High-Enriched Spent Fuel Chemical Processing.................................................................. 8

5.1.3 Other Nuclear Facilities and Processes ...............................................................................9

5.1.4 Radioactive Nuclides of Concern and Solubility ................................................................... 9

5.2 INEEL Radiological Protection Program Management and Support.......................................... 11

5.2.1 Internal Dosimetry Issues Related to Contractor Changes................................................. 11

5.2.2 Bioassay Programs............................................................................................................ 12

5.2.3 Internal Dose Records ....................................................................................................... 12

5.2.4 Internal Dose Regulations, Investigation Levels, and Data Codes ..................................... 15

5.3 Internal Dose Control ................................................................................................................18

5.3.1 Air Monitoring .................................................................................................................... 19

5.3.2 Early Technical and Analytical Capabilities at NRTS ......................................................... 20

5.4 Minimum Detectable Activities .................................................................................................. 22

5.5 Development and Impact of Whole Body Counting ................................................................... 22

5.6 Specific Facilities ......................................................................................................................27

5.6.1 Test Area North .................................................................................................................27

5.6.1.1 Specific Manufacturing Capability Project ................................................................. 28

5.6.2 Idaho Chemical Processing Plant (ICPP) – currently INTEC.............................................. 29
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5.6.2.1 High-Enriched Spent Fuel Storage............................................................................ 30

5.6.2.2 High-Level Wastes.................................................................................................... 31

5.6.2.3 High-Level Waste Calcination ................................................................................... 31

5.6.2.4 Process Analytical Facilities ...................................................................................... 31

5.6.2.5 Spent Fuel Processing and ICPP Most Limiting Radionuclides ................................. 31

5.6.3 Argonne National Laboratories-West ................................................................................. 33

5.6.4 Radioactive Waste Management Complex ........................................................................ 33

5.6.5 Waste Reduction Operations Complex .............................................................................. 35

5.6.6 Test Reactor Area.............................................................................................................. 35

5.6.6.1 TRA Reactors ........................................................................................................... 35

5.6.6.2 TRA Laboratories...................................................................................................... 37

5.7 Default for Missed Dose............................................................................................................ 37

5.8 Unmonitored Workers ...............................................................................................................39

References ........................................................................................................................................40

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

LIST OF TABLES

Table

5.1.4-1 INEEL recorded internal doses for 1992-2000.................................................................. 

Page

10

5.1.4-2 Primary isotopes of concern at the INEEL ........................................................................ 11

5.2.2-1 Routine bioassay history summary................................................................................... 13

5.2.4-1 Internal dose assessment information post 1989.............................................................. 16

5.2.4-2 Internal dose assessment information pre 1989. .............................................................. 16

5.2.4-3 Analytical information that may be in claimant dose files. ................................................. 17

5.2.4-4 Area codes that may be in claimant dose files.................................................................. 17

5.2.4-5 Derived investigation levels (μCi) in 1977 for acute exposures......................................... 18

5.3-1 Surface Contamination Control and MDAs. ...................................................................... 19

5.3.2-1 Urinalysis results in 1959, 1960, and 1961 ....................................................................... 21

5.3.2-2 Detection limits applicable to environmental sample analyses (1953 to 1965).................. 22

5.4-1 Internal dosimetry in vitro MDAs for urine samples by date .............................................. 23

5.4-2 Internal dosimetry in vitro MDAs for fecal samples by date............................................... 24

5.4-3 Internal dosimetry in vivo MDAs by date........................................................................... 25

5.5-1 Summary statistics from the 1963 whole-body counting program..................................... 27

5.6.1-1 Mass and activity ratios of SMC DU uranium isotopes ..................................................... 28

5.6.2-1 Notable Airborne Incidents at ICPP.................................................................................. 30

5.6.2.5-1 Radiologically significant radionuclides for ICPP processed fuels..................................... 32

5.6.3-1 Radionuclides of concern, ANL-W locations, and MDAs................................................... 34

5.6.4-1 Radioactive waste inventory in the TSA ........................................................................... 34

5.6.4-2 Radioactive waste inventories in the active pits in the SDA.............................................. 35

5.6.6.1-1 Radiologically significant radionuclides for ATR fuel gaseous releases ............................ 36

5.6.6.1-2 Major airborne incidents at TRA ....................................................................................... 37

5.7-1 Default table for missed dose ........................................................................................... 38
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 08/18/2003 00-A New Technical Basis Document for the Idaho 

National Engineering and Environmental 

Laboratory (INEEL) – Occupational Internal Dose. 

Initiated by Norman Rohrig.

Draft 12/04/2003 00-B Incorporates internal changes. Initiated by 

Norman Rohrig.

Draft 02/02/2004 00-C Incorporates additional internal changes and 

mandatory paragraphs. Initiated by Norman 

Rohrig.

Draft 07/29/2004 00-D Incorporates internal review comments. Initiated 

by Norman Rohrig.

Draft 08/06/2004 00-E Incorporates NIOSH review comments. Initiated 

by Norman Rohrig.

Draft 08/11/2004 00-F Incorporates NIOSH review comments. Initiated 

by Norman Rohrig.

08/11/2004 08/11/2004 00 First approved issue. Initiated by Norman Rohrig.
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-5 Page 4 of 49

ACRONYMS AND ABBREVIATIONS 

α alpha particle

ACGIH American Conference of Governmental Industrial Hygienists

AEC Atomic Energy Commission

AECL Administrative Exposure Control Level

AEDE Annual Effective Dose Equivalent

AL Analytical Laboratory

AMAD activity median aerodynamic diameter

AMWTP Advanced Mixed Waste Treatment Program

ANL Argonne National Laboratory

ANL-W Argonne National Laboratory West

ANP Aircraft Nuclear Propulsion

ANPP Aircraft Nuclear Propulsion Program

ANSI American National Standards Institute

ATR Advanced Test Reactor

β beta particle 

Bq becquerel

BZ breathing zone (air monitor)

CAM continuous air monitor

cc cubic centimeter

CEDE committed effective dose equivalent

CERT Controlled Environmental Radioiodine Tests or 

Controlled Environmental Radionuclides Tests

CFA Central Facilities Area

CFR Code of Federal Regulations

CHTRU contact handled TRU

Ci curie

COO Chicago Operations Office

CPP Chemical Processing Plant

cpm counts per minute

d day

D&D decontamination and decommissioning

DAC Derived Air Concentration

DAC-hr Derived Air Concentration hour 

DOD U.S. Department of Defense

DOE U.S. Department of Energy

DU depleted uranium

DOE-ID DOE-Idaho Operations Office

DOELAP DOE Laboratory Accreditation Program

dpm disintegrations per minute

EBR Experimental Breeder Reactor

EBR-I Experimental Breeder Reactor No.1

EBR-II Experimental Breeder Reactor No.2

ERDA Energy Research and Development Administration

ETR Engineering Test Reactor

F Fast Absorption Type
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FAST Fluorinel Dissolution Process and Fuel Storage Facility

FCF Fuel Cutting Facility

γ gamma

h hour

HFEF Hot Fuel Examination Facility

HP health physicist/health physics

H&S Health and Safety

HSD Health and Safety Division

HSL Health Services Laboratory

ICPP Idaho Chemical Processing Plant

ICRP International Commission on Radiological Protection

IDO Idaho Operations Office

IET initial engine test

IMBA Integrated Modules for Bioassay Analysis

INC Idaho Nuclear Corporation

INEL Idaho National Engineering Laboratory

INEEL Idaho National Engineering and Environmental Laboratory

INTEC Idaho Nuclear Technology and Engineering Center

L liter

LANL Los Alamos National Laboratory

LLW low level waste

keV kiloelectron volt, 1000 electron volts

μCi/cc microcurie per cubic centimeter

μg microgram

μm micrometer

m meter

M moderate absorption type

MAP mixed activation products

MFP mixed fission products

mCi millicurie

MeV megaelectron volt

mg milligram

MDA Minimum Detectable Activity or Amount

MDL Minimum Detectable Level

min minute

mL milliliter

M & O Management and Operating (Contractor)

MPBB Maximum Permissible Body Burden

MPLB Maximum Permissible Lung Burden

MPC Maximum Permissible Concentration

MPCa MPC for airborne activity

mrem millirem

MRL Minimum Recording/Reporting Level

MTR Materials Testing Reactor

NaK sodium potassium
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NaI(Tl) sodium iodide doped with thallium

NBS National Bureau of Standards

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH OCAS Claims Tracking System

NRC Nuclear Regulatory Commission

NRF Naval Reactors Facility

NRTS National Reactor Testing Station

OCAS (NIOSH) Office of Compensation Analysis and Support

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory 

PCS primary coolant system

PBF Power Burst Facility

pCi picocurie

QA quality assurance

RAF Remote Analytical Facility

RaLa radioactive lanthanum

RAM Radiation/Remote Area Monitor

RBE Relative Biological Effectiveness

RCIMS Radiation Control Information Management System

RDR Radiation Dosimetry and Records

RDL Research and Development Laboratories

RESL Radiological Environmental Sciences Laboratory

rep roentgen equivalent physical

RU recycled uranium

RWMC Radioactive Waste Management Complex

S slow absorption type

SDA Subsurface Disposal Area

SL-1 Stationary Low-Power Reactor

SMC Specific Manufacturing Capability

Τ1/2 Half-life

TAN Test Area North

TBD technical basis document

TLV Threshold Limit Value

TMI Three Mile Island

TRA Test Reactor Area

TSA Transuranic Storage Area

TRU transuranic

WBC whole body counter or whole body count

WERF Waste Experimental Reduction Facility

WIPP Waste Isolation Pilot Plant

WROC Waste Reduction Operations Complex

ZPPR Zero Power Plutonium Reactor (later Zero Power Physics Reactor)
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5.1 INTRODUCTION AND HISTORICAL OVERVIEW

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

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

They will be revised in the event additional relevant information is obtained about the affected site(s). 

These documents may be used to assist NIOSH in the completion of the individual work required for 

each dose reconstruction.

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

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

employer facility” or a “Department of Energy facility” as defined in the Energy Employee 

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).

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, which had a marked influence on the conduct of the internal dosimetry programs at each of the 

facilities are listed as follows:

• The original mission of the NRTS was (as the name implied) high enriched (>50% and mostly 

>90%) uranium reactor concept development, materials testing through high flux test reactor 

operation, and chemical processing of those high enriched (valuable) uranium fuels. The 

production of weapons grade nuclear materials was not a mission.

• The NRTS beginning was 8-10 years later than Oak Ridge National Laboratory and Hanford. 

During those developmental years significant technical progress in professional skills, 

instrumentation, analyses, procedures, and techniques were accomplished. Radiation safety 

programs and techniques from Oak Ridge (ACC 1952) were adopted at startup of the NRTS 

facilities. 

• Two AEC Field Offices (Chicago and Idaho) were responsible for and had oversight of INEEL 

programs included in this report. In addition the Nuclear Navy under direction of the Pittsburgh 

Field Office administered programs and used facilities on the NRTS for training and program 

development, making three Federal organizations utilizing the NRTS and its infrastructure. 

Personnel employed in the Naval program and site are not included in this study.

• In addition to the Federal Agencies involved at the Site, there were numerous contractors through 

the years, who operated the many facilities for the Federal Agencies, and shared some support 

personnel to varying degrees.

• In order to provide needed consistency of radiation safety programs at the NRTS among a large 

variety of facilities and constantly changing Contractors, the AEC established a unique feature at 

the NRTS of 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) technical support in quality 
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assurance of external and internal radiation dose evaluation, 6) maintenance and documentation 

of formal/legal personnel dosimetry records, and 7) research and development in these areas of 

responsibility.

• Though the design and administration of the radiation safety programs in the work place were the 

responsibility of each facility contractor, oversight of the programs was conducted by the AEC; and 

technical data, information (particularly in the instances of detectable worker intake), and 

analytical internal dose calculations/evaluations were exchanged between the AEC HSL and each 

contractor.

As a consequence, and in spite of the constant changes at the NRTS/INEEL, basic assumptions 

related to minimum detectable activities (MDAs) or minimum detectable levels (MDLs), missed dose 

potential, etc. are relatively consistent across the INEEL through the years. There were differences in 

the available nuclear materials from facility to facility, however, as early as 1955-6 gamma spectral 

analysis capabilities at the NRTS 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 identity through gamma spectral and/or 

chemical separation. If the bioassay records do not include the radionuclide analyses and only record 

gross beta or alpha results, default assumptions are described in the following text.

There were times when incidents occurred. It was the policy to thoroughly investigate incidents and 

identity all individuals involved in the incident. When there is no evidence in the incident file or the 

individual’s dosimetry file that an individual was involved, and no other supporting evidence 

supporting that an individual was involved in the incident, it should be assumed that the individual was 

not involved in the incident.

The facility descriptions below are provided to aid the internal dose reconstruction through increased 

general understanding, data interpretation, defaults, etc. Specific information necessary for dose 

reconstruction is contained in Section 5.5.

5.1.1

NRTS, which was the primary nuclear reactor development laboratory in the United States, tested or

evaluated more than 100 reactor concepts (DOE 1997). Fifty-two test reactors have been designed, 

constructed, and operated (including operation-to-destruction tests) on the NRTS. INEEL has 

experienced a number of reactor episodic events, both planned and accidental [for example, the 

Stationary Low-Power (SL-1) military reactor accident on January 3, 1961 (Stacy 2000); a series of 

deliberate safety experiments, conducted by Argonne National Laboratory West (ANL-W), in which 

reactors were allowed to go “prompt-critical” with resultant reactor destruction (Stacy 2000); and the 

Aircraft Nuclear Propulsion (ANP) project operated initial engine tests (IET) with large environmental 

releases in the 1950s (DOE 1991)]. External and internal doses to workers, both expected and 

accidental, were associated with these events (RAC 2002) 

Test Reactors

The largest internal exposures at INEEL have resulted from accidental intakes associated with 

episodic events or planned major releases, for which the time and characterization of the materials of 

the intakes were well known. These exposures were documented in each exposed employee’s file.

5.1.2

In addition to being the site for experimental test reactors, INEEL is the home of the Idaho Chemical 

Processing Plant (ICPP), currently known as the Idaho Nuclear Technology and Engineering Center 

(INTEC). The ICPP reprocessed high-enriched reactor fuel (235U enrichments in the expended fuel of 

High-Enriched Spent Fuel Chemical Processing
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50% to 93%) for 39 years, beginning in 1952 and ending in 1991. Aged mixed fission products 

(MFPs) were the predominant internal hazard, although enriched uranium isotopes and plutonium 

isotopes (Pu-238 enhanced) were limiting in specific process locations. ICPP experienced not only 

operational containment barrier failures but also had accidental criticality events in 1959, 1961, and 

1978. Because the criticality accidents occurred in process vessels in heavily shielded cells, these 

events resulted in relatively minor worker intakes (Stacy 2000, AEC 1960, AEC 1962a). These 

exposures are also documented in the personnel dosimetry files. 

5.1.3

Other nuclear facilities on the INEEL that resulted in internal dose potential or experience include 

those described below.

Other Nuclear Facilities and Processes

The Radioactive Waste Management Complex (RWMC) handled radioactive wastes generated by 

nuclear facilities on INEEL and was the primary disposal location for the Rocky Flats Plant. Although 

most waste came to the RWMC in packages, accidents have occurred during handling and

processing, resulting in releases, this in turn caused intakes of aged MFP, U isotopes, transuranic 

(TRU) radionuclides, and aged mixed activation products (MAP) (Reilly 2001).

The Specific Manufacturing Capability Project (SMC) is a depleted uranium specialty-parts production 

plant that was built in 1985 in the ANP hanger on the Test Area North (TAN) site. The SMC 

processes metric tons of depleted U metal for the production of military shielding units (Stacy 2000). 

The processes of cutting, machining, and handling U metal produce environments in which both 

chronic and accidental intakes of depleted U have occurred.

At the Naval Reactors Facility (NRF), the U.S. Navy used the INEEL for operating reactors and as a 

naval reactor training center. Because this is not a U.S. Department of Energy (DOE) program and 

not under the oversight of DOE, NRF is not part of the dose reconstruction and compensation 

program. However, through the years NRF has participated in limited coordination of radiological 

protection programs and Site support services. It is possible that some workers’ internal dose could 

have resulted from their support work at the NRF.

5.1.4

INEEL facilities and activities have been related primarily to experimental reactor design and 

development, spent fuel processing, depleted uranium parts production, and low- and high-level 

radioactive waste treatment and disposal and are more fully described in the INEEL Site Description 

TBD (ORAU 2003). Table 5.1.4-1, which lists the internal doses received at the INEEL from 1992 to 

2000, demonstrates the radionuclides of more recent concern from an internal dose standpoint are 

MFP (primarily aged), uranium and decay products, and TRU and decay products. This information 

was presented in Reilly 2001 and can be useful to the reconstruction effort primarily as an indicator of 

intake radionuclide identity and quantities for this recent ten-year period. Prior to 1990 higher doses 

were received from the radionuclides listed above, as well as MAP (primarily longer-lived, i.e., Ag110m and others). Most internal doses have been identified following an incident rather than as a 

result of routine bioassay measurements.

Radioactive Nuclides of Concern and Solubility

From these programs and as documented in INEEL Management and Operating (M&O) Contractor 

Technical Basis Document (TBD) for Internal Dosimetry, General Technical Basis and Facility Specific 

Documents (Reilly 2001) and from Health Services Laboratory/ Radiological Environmental Sciences 

Laboratory (HSL/RESL) reports, Table 5.1.4-2 lists radionuclides of concern at INEEL. These 

radionuclides are those for which internal dose has been determined in the past and/or methods have 

been developed for their detection. The INEEL program follows the DOE guidance and uses the 
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ICRP 30 nomenclature for solubility as noted in the table. Also provided is the ICRP 68 (ICRP 95)

recommended absorption type.

Table 5.1.4-1. INEEL recorded internal doses for 1992-2000. 

Year of dose 

assignment Facility CEDE (mrem) Radionuclides of concern

2000 SMC 10, 31, 20, 10, 13, and 20 U-238

2000 INTEC (ICPP) 11 Pu-238, Pu-239/240

1999 SMC 15, 48, 13, and 12 U-238

1999 INTEC (ICPP) 16 Pu-239/240

1998 SMC 16 U-238

1997 TRA 10a and 10 a Eu-152 and Eu-154

1997 INTEC (ICPP) 24 Pu-238 and Pu-239/240

1997 TAN 13 a Am-241, Cs-137, Sr-90, U-233, U-238, Pu238, Pu-239/240

1997 SMC 16 & 20 U-238

1996 RWMC 43 Pu-239

1996 INTEC (ICPP) 15, 87, 136, 652 a , 655 a , 677 a and 678 a Am-241, Cs-137, Sr-90, Pu-238, Pu-239/240

1996 SMC 10, 10, 12, 16, 17, 18, 20, 20, & 23 Am-241, Cs-137, Sr-90, PU-238, Pu-239/240

1995 INTEC (ICPP) 10, 13, 14, 15, 23, 28, 29, 42, 45, & 53 Am-241, Cs-137, Sr-90, Pu-238, Pu-239/240

1995 SMC 10, 12, 14, 15, 19, & 26 U-238

1994 INTEC (ICPP) 14, 20, 25, & 29 Am-241, Cs-137, Pu-238, Pu-239/240

1994 SMC 10, 10, 12, & 15 U-238

1993 INTEC (ICPP) 14, 35, 36, 39, 50, & 53 Am-241, Pu-238, Pu-239/240

1993 SMC 11 U-238

1992 RWMC 20 & 205 Am-241, Pu-238, Pu-239/240

1992 SMC 11, 12, 12, 14, 15, 15, 16, 20, 32, & 52 U-238 a Internal doses determined after a “significant” event when intake occurred. However, even the other intakes were defined 

as a result of known monitoring “flags” being identified.

Through the years at INEEL, plotting urine and fecal elimination curves has shown that an “effective 

solubility” of M to S per the ICRP 66 system (ICRP 94) is to be expected for the radionuclides listed in 

Table 5.1.4-2, except for Sr, Cs, I, Hg, and tritium. Elimination curves for 3

H, 90Sr, 131I, 137Cs, and 

203Hg show them to be typically of F type absorption. The chemical explanation is that radioactive 

materials in oxygen atmospheres oxidize rapidly, forming less soluble compounds. The default 

assumption of M or S would be appropriate, based on the most claimant-favorable result to the organ 

in question. 
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Table 5.1.4-2. Primary radionuclides of concern at INEEL. 

Element Radionuclides

ICRP 30a

Retention 

Class

ICRP 68 

Absorption 

Type

Preferred 

analysis technique

hydrogen H-3 (assume HTO) D F Urine

chromium Cr-51 F,M,S WBC

manganese Mn-54 F,M WBC

iron Fe-59 F,M WBC

cobalt Co-58, Co-60 W,Y M,S WBC

zinc Zn-65 S WBC

strontium Sr-89, Sr-90/Y-90b D F Urine

zirconium/niobium Zr/Nb-95 W,Y F,M,S WBC

molybdenum Mo-99 F,S Urine

technetium Tc-99 F,M Urine

ruthenium Ru-103, Ru-106 F,M,S Urine

silver Ag-110m F,M,S WBC

antimony Sb-122, Sb-125 F,M WBC

tellurium Te-132 F,M Urine, WBC

iodine I-131, I-133 (assume elemental I2) F Urine, thyroid counter

cesium Cs-134, Cs-137 D F WBC

barium/lanthanum Ba/La-140 F/F,M WBC

cerium Ce-141, Ce-144 W,Y M,S WBC

europium Eu-152, Eu-154, Eu-155 W,Y M WBC

gadolinium Gd-153 F,M WBC

tantalum Ta-182 M,S WBC

mercury Hg-203 (assume inorganic) F Urine, WBC

protactinium Pa-233 M, S WBC

uranium U-233, U-234, U-235, U-236, U-238 D,W,Y F,M, S Urine, fecal

neptunium Np-237 M Urine, fecal

plutonium Pu-238, Pu-239/240 W,Y M, S Urine, fecal

americium Am-241 W M Urine, fecal, lung 

counter a Assigned in the INEEL internal dosimetry technical basis document (Reilly 2001) 

b Y has an absorption type of M or S; however in the case of short-lived Y-90, it tracks its parent Sr-90 which is 

normally a type F at the INEEL.

5.2 INEEL RADIOLOGICAL PROTECTION PROGRAM MANAGEMENT AND SUPPORT

5.2.1

The changes in contractors at INEEL during its 54-year history are listed in Tables 2.1 and 2.2 of the 

INEEL Site Description TBD of this Site Profile (ORAU 2003) and resulted in relatively frequent

management changes at most of the facilities. The contract with the University of Chicago to operate 

the ANL-W facilities has not changed in this 54-year period. 

Internal Dosimetry Issues Related to Contractor Changes 

The primary AEC/DOE oversight for NRTS/INEEL, which included most projects on the Site and all 

support functions, was assigned to the AEC Idaho Operations Office (IDO). A unique AEC function 

organized as the Health and Safety (H&S) Laboratory provided a variety of health and safety support 

functions to the entire Site, which included external and internal dosimetry, health physics 

instrumentation, fire department, medical department, and environmental surveillance. The Chicago 

Operations Office (COO) provided oversight for ANL-W (contracted by the University of Chicago) 

programs and facilities, while the U.S. Navy provided oversight for the NRF program. ANL-W, which 
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is a DOE program, is included in this TBD. The ANL-W program used Site support services, including 

internal dosimetry support, although the bioassay results will be reported through COO. 

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

by the AEC Health and Safety Laboratory and the subsequent managers of these support functions. 

Records related to the individual facility/contractor field monitoring programs (air monitoring data, 

personnel contamination records, etc.) were maintained by individual contractors and/or site areas 

and are not maintained in a single recordkeeping system. The field monitoring data were not 

available for use in this report. 

In spite of the constant change in operational responsibility through the years and the movement of 

workers among facilities, there has been a basic level of site-wide consistency in the internal 

dosimetry programs applied to INEEL facilities and programs, and particularly the bioassay analytical 

techniques and calculational processes. Though the field programs (which monitored the workplace 

and identified work groups to be included in the routine bioassay programs and workers, who needed 

special bioassays) were implemented by the individual contractors, there was routine interaction with 

the H & S Laboratory professionals in interpretation of dosimetry results, as well as in determination of 

any needed corrective practices/procedures. 

Typically, employees were assigned to individual facilities and were monitored for specific radiological 

hazards associated with the work. During periods when a single prime contractor was responsible for 

programs at most facilities or for site-wide support personnel, workers in certain crafts [e.g., 

maintenance, specialty operators, and some health physics (HP) technicians] worked at multiple 

facilities and were exposed to a variety of radioactive materials in a variety of work situations. 

The internal dose reconstruction for personnel who have worked at a number of INEEL facilities 

should rely on specific bioassay data (radionuclides, quantities, etc.) when available. The procedures 

and technical capabilities for collecting and analyzing bioassay samples at the different facilities were 

basically equivalent. In addition, both the individual facilities and the AEC H&S Laboratories had 

radionuclide identification capabilities from the early 1960 time period. Positive bioassay results 

(analyses in which the results exceeded 2-sigma counting statistics) were normally followed by a 

confirmatory analysis, identifying a specific radionuclide. In the cases where only gross beta and/or 

alpha bioassay results are available, the results are normally less than two-sigma. However if it is 

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

defaults should be Sr/Y-90 and Pu-238, respectively. 

5.2.2

Routine bioassay of radiation workers has been conducted since the beginning of the site. However, 

formal documentation of the bioassay programs has not been found for times before 1981. Some of 

the data sheets on individuals indicate that bioassay sampling was routinely done every 6 months in 

1953. Table 5.2.2-1 presents the reconstructed history of routine bioassay frequency. 

Bioassay Programs

5.2.3

Formal or “legal” internal dose data were maintained by the DOE HSD in individual hard-copy folders 

until 1989, 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 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 current contractor subject matter expert reviews, validates, 

and/or prepares official internal dose assessments. A DOE staff member at RESL is responsible for 

Internal Dose Records
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-5 Page 13 of 49

oversight of INEEL internal dosimetry program functions and provides quality assurance. The RDR 

unit function includes documentation and records custodial responsibilities. In 1999, the Radiation 

Control Information Management System (RCIMS) was placed in service to support the Radiation 

Protection program, including internal dosimetry. RCIMS is an electronic database that carries 

reported internal doses as committed effective dose equivalent when an individual’s dose history is 

prepared.

Table 5.2.2-1 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 

5.3.4-1.

AEC 1958

AEC 1961

1961 Annually in vitro

 urine

in vivo

Radiation workers Frequency is 

inferred from 

individual data 

sheets.

Table 

5.3.4-1.

AEC 1962a

1962- 

1972

Annually in vitro

 urine

in vivo

Radiation workers Frequency is 

inferred from 

individual data 

sheets.

AEC 1962a

AEC 1963b

1973 - 

1981

Annually

When 

internal 

intake 

suspected

Annually

in vitro

urine

fecal

in vivo

Radiation workers

Annual DE >10% 

quarterly standard in 

ERDA Manual Chapter 

0524.

Reporting Frequency is 

inferred from 

individual data 

sheets.

AEC 1968 

AEC 1975

ERDA 1975

1982-

1987

Annual

When 

internal 

intake 

suspected

Annual

in vitro

urine 

fecal

in vivo

CPP-603 workers

Fuel reprocessing 

operators

Waste reprocessing 

operators

Shift lab workers

Health physics 

technicians

Selected radio-chemistry 

workers 

Maintenance workers

Denitrator operators

50-year CDE >10% 

quarterly standard in 

ERDA Manual Chapter 

0524.

Reporting Staggered to 

monitor group 

throughout the 

year

Exxon 1981

termination in vivo All radiation workers

1988-

1989

1 to 6 

months

in vivo All radiation workers

Lung 50-yr CEDE > 

0.5 – 1.0 rem

Investigating

Bone Surface 50-yr 

CEDE > 1.0 – 2.0 rem

Other Organs 50-yr 

CEDE > 0.5 - 1.0 rem

Staggered so 

that worker 

receives some 

sort of analysis 

or sampling 

every 3 

months.

Tschaeche 

1988

Annually in vitro

 fecal

18 to 24 

months

in vitro

 urine

termination in vivo When internal exposure 

in vitro suspected.

new hire in vitro Depending on review of 

in vivo radiation dose history.
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Table 5.2.2-1 Routine bioassay history summary (continued)

Year Typical 

Frequency

Type Groups 

Analyzed/Sampled

Investigating Level Comments Reference

1990-

1994

annually in vitro All radiation workers 

where exposure to 

surface or airborne 

radioactive contamination 

could give at least 0.1 

mrem AEDE from 

occupational sources, or 

give an organ or tissue 

DE > 5 rem annual.

In accordance with 

DOE Order 5480.11. 

Reporting

Workers that could 

receive 0.1 rem AEDE 

or 5 rem AEDE organ or 

tissue dose.

AEDE ≥ 0.01 rem.

Investigating

Bioassay 

requested

when 

workplace 

monitoring 

program 

indicates > 

0.02 ALI.

Follow-up 

triggered by 

positive results 

from the 

workplace 

monitoring 

program, 

positive routine 

bioassay 

sample, or in 

response to 

incidents 

involving 

suspected 

intakes.

King 1990

6 months in vitro Rich 1990

 fecal

annually in vitro

 urine

new hire In vivo Worked at a facility where 

gamma emitting 

radionuclides were 

handled.

in vitro

 urine

 fecal

Worked in U mining, 

smelting, manufacturing, 

recovery, or enriching 

facilities.

Worked with transuranics

termination in vivo

in vitro

Any employee suspected 

of having an internal 

exposure or that was on a 

scheduled monitoring 

program.

1995 Appropriate 

to the 

facility 

mission, 

potential 

uptakes.

When 

work-place 

monitoring 

indicates 

significant 

potential 

for intakes.

in vivo 

in vitro

 urine

 fecal

All radiation workers that 

enter radiological buffer 

areas or areas of higher 

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 U and TRU source 

terms.

Random sampling is 

performed to demonstrate 

the adequacy of the 

radiological controls in 

limiting the internal intake 

of radionuclides. 

Employees are selected 

at random form both nonradiation workers and a 

radiation worker 

population.

In accordance with 

DOE 5480.1 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. 

Internal doses resulting 

from all confirmed 

intakes are to be 

evaluated.

Investigating

Each facility 

has a specific 

Technical Basis 

Document for 

Internal 

Dosimetry.

Follow-up for 

any suspected

intake of 

radionuclides 

and to more 

accurately 

identify and 

characterize 

the amount of 

intake and 

excretion 

pattern.

Andersen 

1995
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Table 5.2.2-1 Routine bioassay history summary (continued)

Year Typical 

Frequency

Type Groups 

Analyzed/Sampled

Investigating Level Comments Reference

1995-

2000 

new hire Based on screening to 

determine internal 

conditions from previous 

uptakes or to establish 

baseline for those 

continuing to work s 

radiation workers.

termination in vivo

in vitro

Any employee that was on 

a scheduled monitoring 

program.

2001 As 

developed 

by individual 

facilities 

based on 

analysis 

tables 

developed 

for each 

radionuclide.

in vivo 

in vitro

 urine

 fecal

All radiation workers

In accordance with 

DEOE 5480.1 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.

Investigating

uranium > 1.0 μg/l

 

in vitro activity 

detected > 2σ. 

in vivo > 2.33 σ. 

 Default trigger levels 

exceeded. 

Bioassay is 

mandatory 

when an 

employee or 

visitor is 

involved in an 

event where the 

internal uptake 

of radionuclides 

was likely to 

have occurred.

Rielly 2001

termination in vivo

in vitro

Any employee that was on 

a scheduled monitoring 

program.

5.2.4

The following information is important to internal dose reconstruction, since the claimant files from 

DOE can contain a variety of internal dose information, ranging from the calculated internal doses as 

well as the in vitro and in vivo individual bioassay results. The 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.

Internal Dose Regulations, Investigation Levels, and Data Codes

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

years. Table 5.2.4-1 contains the description of coded information that may be encountered post 

1989. Table 5.2.4-2 contains internal dose information that may be in pre-1989 records. Table 5.2.4-

3 contains analytical nomenclature that may be encountered. Table 5.2.4-4 contains INEEL area 

codes.

Federal regulations related to permissible internal dose and formal reporting requirements to the 

AEC/ERDA/DOE changed periodically through the years of the history of the NRTS/INEEL. While the 

regulations influenced the level of calculated internal dose, which would be reported and/or 

documented, these changes did not significantly change the analytical programs for the detection of 

internal intake. For example, the monitoring and analytical programs were designed to initiate an 

investigation of any potential internal intake as indicated by positive air sampling, personnel 

contamination, etc., by in vitro and/or in vivo bioassay analyses. Most of these recorded analyses did 

not result in detectable radionuclides. During the early years internal dose was usually considered 

separately from external dose in terms of meeting specific exposure limits and the calculated dose 

was only reported and documented if specific dose levels were exceeded (AEC/ERDA 0524, 1968-
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1977, required periodic urinalyses and/or in vivo counting and/or evaluation of air concentrations if the 

whole body dose or dose commitment could exceed 300 mrem in a calendar quarter). Changes in the 

reporting levels did not generally result in changes to the air monitoring and bioassay sampling 

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

regardless of the formal reporting requirements. 

Table 5.2.4-1 Internal dose assessment information post 1989. 

Coded 

Information

Description

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

Asmt. Nos. This assessment number is the calendar year, i.e., (83)- No. 

represents 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 30 solubility 

class symbol D, W, or Y. Amount in either μCi or Bq

CEDE rem Calculated Committed Effective Dose Equivalent in rem

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

Organ CDE rem Committed Dose Equivalent calculated for the listed organ in rem

Employer and Exp. 

Location

Abbreviation of DOE Site Contractor and the plant site of exposure 

(may 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 5.2.4-2 Internal dose assessment information pre 1989. 

Dose

Information

Description

Name, Soc. Sec. 

No.

Employee name, social security number, and (Contractor 

Abbreviation/Plant or Facility)

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

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

year in which exposure occurred

CEDE rem Calculated Committed Effective Dose Equivalent in rem

AEDE rem Calculated Annual Effective Dose Equivalent in rem

Year Year for which the AEDE was calculated

The “investigation levels” have also changed little from the early times to the present, i.e. positive 

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

body. Dickson (1977) established official investigation levels, which are shown in Table 5.2.4-5, for 

acute uptakes of radionuclides corresponding to one-tenth of the quarterly radiation standard. Later 

procedures (DOE 1988) have set specific limits on those positive bioassay results which could result 

in 100 mrem annual EDE or above as the point at which followup and reporting was required. With 

the DOE Radiological Control Manual, this changed to 100 mrem CEDE. Also 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 non-detectable radioactivity in 

bioassay samples, although the request for and number of followup samples and/or analyses may 

have been different as a function of the formal regulations in effect.
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Table 5.2.4-3 Analytical information that may 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 that were performed, i.e. 

Sr-90, Y Separation, etc.

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

evaporated for gross beta and/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 K and/or U

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 – c/m Counts/min. determined by dividing total counts by time of count

Bkgd. c/m Background counts/minute recorded

Net count-c/m Gross c/m minus background c/m

K40 corr. – c/m Additional background from K-40 identified. 40K is not a facility 

occupational product – ignore from an internal dose reconstruction

Foreign Activity –

c/m and d/m

Net counts corrected for 40K and then converted to d/m based upon 

counter calibration. Uncertainty also included, which is recorded 

as 1 standard deviation 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 5.2.4-4 Area codes that may be in claimant dose files. 

Area Code

Description

Area Code

Description

1 AEC Headquarters Bldg 20 TREAT

2 EBR – I 21 LX

3, 034, 035 CFA 22 GCRE

4, 042,045 MTR, TRA 23 OX

5, 053, 055 Id Chemical Processing Plant 24 ARHG

6 NRF 25 No information available

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

8 Services 27 ML-1

9 NX 28 On-Site Site Survey

10 AX 29 Off-Site Site Survey

11, 113 CX 30 ANP at SL-1

12 EX 31 STPF

13, 133,135 SPERT, PBF 65 ECF

14 MORE 66 Non-Security

15 SX 67 Division of Compliance

16 SL-1 68 STEP

17, 333 MX 69 LPTF (Phillips & AEC)

18, 814,815 WP, RWMC 71 CADRE

19, 772, 775 TAN (Phillips & AEC) 774,776 SMC
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Table 5.2.4-5 Derived investigation levels (μCi) in 1977 for acute exposures. 

Radionuclide

Inhalation

(Lung Burden)

Ingestion

(Total Activity)

 Cr-51 20 500

 Mn-54 0.4 30

 Co-57 2 90

 Co-60 0.09 9

 Zn-65 0.6 30

 Zr-95 0.3 20

 Ru-106 0.06 3 

 Sb-125 0.3 30

 Cs-134 0.1 3

 Cs-137 0.1 4

 Ce-144 0.06 3

 Pu-239 whenever detected

 Am-241 whenever detected

 Sr-90 (bone) when detected by skull counting

 I-131 (thyroid) initial content 0.27

5.3 INTERNAL DOSE CONTROL 

The radiological protection program was established to provide detection of barrier or ventilation 

failure in a timely manner and 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 (respirators, etc.), and requesting bioassay analyses to identify internal intake. As a 

consequence of a consistent AEC/DOE policy to avoid detectable internal exposures, coupled with 

the time and technical complexity of an internal dose evaluation, the general policy at INEEL for 

internal exposures has been preventive in nature. Generally, radiological materials handled at the 

Site were of relatively low volume and mass and of higher activity concentration, as opposed to metric 

tons of materials of low specific activity. The consistent NRTS/INEEL policy and practice was to 

require respiratory protection on jobs when the possibility of generating airborne contamination was 

thought to exist, regardless of the actual measured air or surface contamination. These practices 

influence the assumptions related to dose evaluation in internal dose reconstruction.

As a related note, the contamination control limits for the detection and control of released activity 

beyond the control boundaries were related to instrumentation capabilities and the basic philosophy of 

acceptance of detectable contamination. As a result of increased emphasis on ALARA, some 

reduction in acceptable release levels was also implemented. The contamination control limits for 

alpha on plant surfaces and particularly personnel were always set close to the MDA, such that “any 

detectable” contamination was a signal for preventative and/or follow-up evaluations and actions. 

Beta/gamma MDAs typically were a factor of 5 below the limits. Table 5.3-1 is a summary of control 

limits primarily from the CPP Health Physics Manual, April 1952 (ACC 1952) and current operating 

procedures.
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Table 5.3-1 Surface Contamination Control and MDAs. 

Time 

Period

Surface Location Detection 

Technique

Control Levels MDA - Typical

1952 –

1960s

Plant/Equipment Smears 500 d/m β & 20 d/m α per 

100 cm2 

150 d/m β & 10 d/m α

per 100 cm2

Personal Clothing Portable Survey 

Instruments

1500 d/m β & 500 d/m α

per 100 cm2

1000 d/m β & 500 

d/m α per 100 cm2

Personal Skin Portable Survey 

Instruments

Any detectable reported, 

e.g. 1000 d/m β & 500 d/m 

α per 100 cm2 

1000 d/m β & 500 

d/m α per 100 cm2

Shipments Smears/Portable 

Survey Instrument

500 d/m β & 20 d/m α per 

100 cm2

 – smears 

0.1 mrep/hr β & 500 d/m α

per 100 cm2

150 d/m β & 10 d/m α

per 100 cm2 smears. 

0.01 mrep/hr β & 500 

d/m α per 100 cm2

1970s -

present

Plant/equipment 

Surfaces

Smears 300 d/m β & 20 d/m α per 

100 cm2

30 d/m β & 10 d/m α

per 100 cm2

Personnel Portable Survey 

Instruments

Any detectable reported, 

e.g. 300d/m β & 200 d/m α

per 80-100 cm2 

300d/m β & 200 d/m 

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 from the 

beginning of all facility and program operations in routinely occupied areas. With the exception of the 

SMC (1985) project, the primary contaminant radionuclides by activity were either MFPs or MAPs, 

which were beta-gamma emitters with Maximum Permissible Concentrations/Derived Air 

Concentrations (MPCs/DACs) in the > 10-9 μCi/cc range. TRU and U were available at some of the 

INEEL facilities, but they were nearly always well-“tagged” with beta-gamma activity that allowed 

beta/gamma-detecting CAMs to be used to warn of possible alpha contamination or internal 

exposures.

Air Monitoring 

A 1952 CPP Health Physics Manual describes a CAM and three other air sampling systems. The 

1952 manual required use of a “filter type respirator” when airborne activity exceeded 10-8 μCi/cc for 

beta-gamma activity or 10-11 μCi/cc for alpha activity. An army assault-type mask was invoked when 

levels exceeded this by a factor of 10. Positive air-supply masks were invoked if levels larger by a 

factor of 1000 were experienced (ACC 1952, p IX: 4-1).

The CAM monitoring systems provided “live” or “real-time” air activity evaluations (although it is not 

clear what the set points for alarms were), and fixed air samplers at several locations provided 

retrospective data and an average air concentration of beta-gamma emitters in an area or building. 

The fixed air filter system samples were counted for both beta and alpha activity. Later alpha CAMs 

were provided in select facilities, where alpha contaminants could be controlling. CAMs were 

calibrated, and training programs for HPs were established for interpreting CAM responses for various 

situations, radionuclides, response times, filter accumulation, etc. In the event personnel were 

required to work in an area or building where known air contamination was present, respirators were 

worn to reduce internal contamination intake to levels below detectable amounts.

Generally workers were asked to submit to bioassay whenever they were in an area in which a CAM 

alarmed. In addition, the fixed location and retrospective air sampling system would signal the need 

for bioassay, if elevated air sample results were detected. 
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However, in the absence of bioassay data for a known radiation worker in one of the functioning 

facilities as a default, claimant favorable, unmonitored dose 100 DAC/MPC-hours for Sr/Y-90 or Pu238 could be assumed per year. This is based upon a consistent and standing policy of taking 

confirmatory bioassay for workers exposed to known levels of air activity. However for an MDA for 

chronic activity detection would be < 0.01 MPC (in the range of 10-11 

μCi/cc for beta) x 10 ratio of 

general area to breathing zone concentration x 1000 hours of undetected exposure = approx 100 

DAC-hrs or approximately 2 x 105 pCi intake.

5.3.2

DOE HSL technical reports and annual reports, coupled with facility memos and reports, document 

the analytical detection capability dating from the early 1950s and 1960s. For example, during ANPIET runs in 1956, particulate and liquid caustic filter samples of effluent were analyzed with gamma 

spectroscopy and specific chemical separations of the identified radionuclides (Ebersole 1956). This 

analytical capability to identify radionuclides by their energy spectra was available and used for urine 

and other bioassay samples. Specific separations (strontium, iodine, etc.) were available to quantify 

the radioactive components of a variety of samples of interest.

Early Technical and Analytical Capabilities at NRTS

In the early days a gross beta measurement was made on an evaporated aliquot and/or gamma count 

directly on a liquid sample. Any detectable activity triggered a specific chemical separation analysis – 

generally for Sr. Early analyses for Pu generally were gross alpha counts on a Pu 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- × 3-in. sodium 

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

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

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

“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, p. 59).

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

counting would be effective in all situations except for exposure to pure strontium isotopes. In order to 

guard against this unlikely possibility, the procedure was established of performing a strontium 

analysis for individual “workers at risk” (radiation workers) every two years and/or at termination. 

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 analysis not completed 

unless a detectable count was obtained on the precipitate. A 100-mL sample of urine permitted the 

detection of approximately 8 x 10-8 μCi/mL of Sr-90.

Table 5.3.2-1 reproduces reports of the urinalysis results for 1959, 1960, and 1961 as obtained from 

Table 4 of AEC (1960), Table IV of AEC (1961), and Table XIII of AEC (1962). The practice was to 

perform a gross beta or gamma analysis and identify specific radionuclides if the gross counts 

indicated activity above background levels. The total number of urinalysis in 1959 (11,066) when 

3524 people had radiation badges with 715 receiving external doses above 500 mrem demonstrates 

that people provided urine samples multiple times during the year. 

Table 5.3.2-2 summarizes early detection limits for various environmental water and milk samples 

between 1953 and 1965. These analytical sensitivities are applicable to analysis of body 

fluids/substances for the purpose of bioassay (AEC 1964). 

1 
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Table 5.3.2-1. Urinalysis results in 1959, 1960, and 1961

Nuclide/

Element 

of 

Interest

Type

Activity

Total Number 

Performed

Statistically Significant

Highest Resulta Number Percent

1959 1960 1961 1959 1960 1961 1959 1960 1961 1959 1960 1961

Gross

βb 8,546 8,546 30 65 15 5 0.76 0.18 17

18,820 ± 632 

d/m/5 mL

992± 40

d/m/5 mL 172±16 d/m/mL

Gross

γc 2,433 2,712 9,120 174 129 --d 7.15 4.76 4

35,972 ± 310 

d/m/5mL

19,817±105 d/m/5 

mL

1,900,235±876 

d/m/75 mL

Co-60 β -- -- 1 -- -- 0 -- -- 0 -- --

300±75

d/m/450 mL

Sr-90 β 3 105 3,248 3 0 2 100 0 2

4.12 x 10-2

d/m/mL Insignificant

183±8

d/m/75 mL

Sr-91 β 20 37 2 19 0 -- 95 0 0 388 ± 1.6 Insignificant 4±8 d/m/mL

I

e β -- 9 -- -- 2 -- 22 0 9992±80 d/m/mL --

Cs-137 β -- -- 40 -- -- 0 -- -- 0 -- --

1460±10% 

d/m/1700 mL

Ba-139 α 20 16 0 -- 80 0 --

120 ± 0.8 d/m/mL

0 --

Thf α 7 0 -- 0 0 -- 0 0 -- Insignificant 0 Insignificant

Ug α -- -- 4 -- -- -- -- -- 0 -- -- 10 μg/L

U-233 α 17 3 -- 1 0 -- .06 0 -- 180 ± 4.0 d/m/mL Insignificant --

Pu-239 α 18 0 29 0 0 0 0 0 Insignificant Insignificant 2E-9 μCi/mL

Am-241 α 2 0 -- 0 0 -- 0 0 -- Insignificant 0 --

Totals 11,066 11,352 12,494 278 146 -- 2.51 1.29 4

 

a All except two 131I exposures in 1961 listed under gross gamma activity are less than 10% of the permissible body burden for the radionuclide of interest b If only Gross β analyses are available, the default should be Sr-90. c If only Gross γ analyses are available, the default should be Cs-137. d No data reported.

e I isotope(s) not identified in reference. Assume I-131.

f

Th isotope(s) not identified in reference. Assume Th-228.

g U isotope(s) not identified in reference. Assume U-235.

1 
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Table 5.3.2-2. Detection limits applicable to environmental sample analyses (1953 to 1965) 

Type of sample Radiation Detection limit

Water Alpha 3E-9 μCi/mL

Beta 6E-9 μCi/mLa

Tritium 4E-6 μCi/mL

Milk Iodine-131 10 pCi/literb

Strontium-90 1.5 pCi/liter a

Reduced in October 1962 from 1.5E-7 μCi/mL b

Reduced in September 1962 from 50 pCi/liter.

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

analytical laboratory. The individual analytical data were recorded, as were the specific doses for 

which formal evaluations were required. Prior to 1977, internal doses were assigned if the internal 

dose or dose commitment was greater than 50% of the yearly allowable body or organ doses by DOE 

Manual Chapter 0524, the controlling regulation at the time (Aoki 1979). In 1977, the policy was 

changed to assign internal doses when the committed dose equivalent (CDE) to an organ exceeded 

one-tenth of the quarterly radiation protection standard. Replies to requests for radiation exposure 

history prior to 1979 stated that there was “no positive exposure reported” when the dose was below 

the reporting levels noted above. Because individuals remembered receiving some internal exposure, 

the statement “no positive exposure reported” was determined to be misleading and was changed in 

1977 to “no reportable levels recorded.” In all cases, copies of the bioassay results were placed in the 

individual’s radiation exposure file and should be in the individual’s dose file on NOCTS. 

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 (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 (2- sigma) 

are performed. Current MDAs for urine and fecal sample analysis are listed respectively in Table 5.4-

1 and Table 5.4.2 along with values gleaned from historical documents. By far the majority of the 

urine samples taken at the Site were single voidings, with 24-hr samples used for special sampling 

purposes, i.e. followup samples, primarily to extend the sensitivity. The MDAs listed are those for the 

primary samples. The recommended time periods for the MDA values have been included in the 

table.

The current in vivo MDAs are listed in Table 5.4-3 along with values gleaned from historical 

documents. The recommended time periods for the MDA values have been included in the table. 

5.5 DEVELOPMENT AND IMPACT OF WHOLE BODY COUNTING

Whole body counting was introduced at the INEEL in 1961. As early as 1961 one of the fundamental 

conclusions from the experience at the NRTS with in vivo and in vitro internal dosimetry analytical 

techniques was the fact that a large proportion of the internal exposures to NRTS workers was to 

insoluble materials. Radioactive nuclides (e.g., 125Sb, 110mAg, 65Zn, and 95Zr/Nb) were detected by an 

in vivo count and not 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-hour specimen (AEC 1962a; Sill et al. 1964). Whole-body counting was demonstrated to 

detect activity as low as 0.01 μCi in a 10 minute count. (AEC 1962a) This detection level was several 

orders of magnitude lower than the maximum permissible body burdens for most beta/gamma fission 

and activation products.
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As a consequence, the in vivo counting program was used to count a) all terminating employees, who 

required physicals exams, b) employees who were suspected of having a possible internal intake, and 

c) selected groups of individuals scheduled for semi-routine analyses by Health Physics Supervisors 

(Sommers 1961). In CY 1963 approximately 1650 counts were performed, with only those activities 

>0.1 μCi being quantified further. This level was determined to be less than one-tenth of the 

maximum permissible body burden for most of the gamma-emitting isotopes. Table 5.4-1 contains 

MDLs for internal dosimetry analytical counting by date.

The 1963 Annual Summary Report (AEC 1964) describes the year’s follow-up for the whole-body 

counting program as presented in Table 5.5-1. As shown in the table, many of the individuals were 

counted multiple times. The maximum activity detected provides an upper bound on how large an 

activity might be found in someone in earlier years before the whole body counter was operational. 

Table 5.4-1. Internal dosimetry in vitro MDAs for urine samples by date. 

Radiation/

Radionuclide Year

Urine 

pCi/mL Reference

Gross β Start date -

1953

8.4 Data Sheet

1954-1959 9.3 Ebersole & Flygare 1957

1960-70 3.5 AEC 1961a

1971-present 1 AEC 1972a

Gross γ 1960-1964 3.5 AEC 1961a

1965-present 2.3 Data Sheet

H-3 Start date -

1994

35 AEC 1972a , ACB 1974

1995-present 1.4 Andersen et al. 1995, 

Rielly 2001

Co-60 1963-present 0.01 Rich 1990

Sr-90 1953-1959 9.2 Ebersole 1957

1960-1970 0.2 AEC 1961

1971-1989 0.02 AEC 1972a, ACB 1974

1990-present 0.01 Rich 1990

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 5.E-5 ACB 1974

Np-237 1974-present 0.00005 ACB 1974

U (FP) 1954-1985 14 μg/L

U (KPA) 1985-2001 0.17 μg/L

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

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

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

Pu-238 1974-present 2. E-5 ACB 1974

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

1971-1973 5.E-4 AEC 1972a

1974-1989 2 E-5 ACB 1974

1990-present 6.E-5 Rich 1990

Am-241 1974-1989 7 E-5 ACB 1974

1990-present 2. E-4 Rich 1990

Cm-244 1974-present 1. E-5 ACB 1974

Cf-252 1974-present 1. E-5 ACB 1974 a MDA calculated from inferred 2σ uncertainty. 
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Table 5.4-2. Internal dosimetry in vitro MDAs for fecal samples by date

Radiation/

Radionuclide Year

Fecala

pCi/

Sample Reference

Co-60 1963-present 10 Rich 1990

Sr-90 1963-1994 10 Rich 1990

1995-present 1.9 Andersen et al. 1995, 

Rielly 2001

Cs-134 1963-present 10 Rich 1990

Cs-137 1963-1999 0.01 Rich 1990

2000-present 0.3 PLN-153 2000

Th-230 1974-present 0.0003 ACB 1974

Np-237 1974-present 0.0003 ACB 1974

U-233/234 1970-2002 0.041 Andersen et al. 1995, 

Rielly 2001

2003-present 0.05 Bhatt 2003

U-235 1970-2003 0.038 Andersen et al. 1995, 

Rielly 2001

2003-present 0.09 Bhatt 2003

U-238 1970-1994 0.5 Rich 1990

1995-2002 0.03 Andersen et al. 1995, 

PLN-153 2000, Rielly 2001

2003-present 0.09 Bhatt 2003

Pu-238 1974-1994 0.03 ACB 1974

1995-2002 0.022 Andersen et al. 1995, 

Rielly 2001

2003-present 0.02 Bhatt 2003

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

1974-1994 0.02 ACB 1974

1995-present 0.03 Andersen et al. 1995, 

PLN-153 2000, Rielly 

2001, Bhatt 2002

Am-241 1974-1994 0.07 ACB 1974

1995-2001 0.023 Andersen et al. 1995, 

Rielly 2001

2002-present 0.04 Bhatt 2002

Cm-244 1974-present 0.02 ACB 1974

Cf-252 1974-present 0.02 ACB 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.
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Table 5.4-3. Internal dosimetry in vivo MDAs by date. 

Radiation/

Radionuclide Year

In vivo 

nCi

 In vivo

Count

Time 

(min)

Reference

γ 

Spectrometer

1974 2.E-4 (thyroid) 10 ACB 1974

Cr-51 1962-2000 12 10 Percival 1962a

2001-present 32 5 Rielly 2001

Mn-54 1962-2000 5 10 Martin 1989, Grothaus 1993, 

Andersen et al. 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 1962a

2001-present 2.5 5 Rielly 2001

2001-present 1.1 10 Rielly 2001

Co-60 1962-1970 12 10 Percival 1962a

1971-1988 5 10 AEC 1972a, ACB 1974

1989 7 10 Martin 1989

1990-1992 2 (lung) Rich 1990

1993-2000 7 10 Grothaus 1993, Andersen et 

al. 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 et al. 1995

2001-present 4.9 5 Rielly 2001

2001-present 2 10 Rielly 2001

Sr-90 1968-1977 70 (skull) 10 Voelz 1969a, AEC 1972a, 

ACB 1974

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

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

1989-2000 5 10 Martin 1989, Grothaus 1993, 

Andersen et al. 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 1962a

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

I-131 1962-1989 12 10 Percival 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

I-133 1989-1998 5 10 Martin 1989, Grothaus 1993, 

Andersen et al. 1995

1999-2000 2 (lung) Rich 1990

2001-present 3. 5 Rielly 2001

2001-present 0.96 10 Rielly 2001 a MDA calculated from inferred 2σ uncertainty. 
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Table 5.4-3. Internal dosimetry in vivo MDAs by date (continued).

Radiation/

Radionuclide Year

In vivo 

nCi

 In vivo

Count

Time (min) Reference

Cs-137 1962-1970 12 10 Percival 1962a

1971-1998 5 10 AEC 1972a, ACB 1974, Martin 

1989, Grothaus 1993, 

Andersen et al. 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 et al. 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 et al. 1995

Ta-182 1962-present 12 10 Percival 1962a

Hg-203 1962-present 12 10 Percival 1962a

Th-230 1974-present 1000 ACB 1974

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

Np-237 1974-present ACB 1974

Uranium 1974 30 mg (lung) 40 ACB 1974

U-233/234 1990-present 0.2 (lung) Rich 1990

U-235 1993 0.2 (wound) 20 Grothaus 1993

1962-present 0.2 (lung) Rich 1990

2001 0.11 (lung) 60 Rielly 2001

U-238 1990-present 0.2 (lung) Rich 1990

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

1989-1998 26 (lung) 60 Martin 1989, Grothaus 1993, 

Andersen et al. 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 1972b

1974-1988 74 (lung) 100 ACB 1974

1989-1988 80 (lung) 60 Martin 1989, Grothaus 1993, 

Andersen et al. 1995

1993 2 (wound) 20 Grothaus 1993

1990-2000 30 (lung) Rich 1990

2001-present 140 (lung) 60 Rielly 2001
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Table 5.4-3. Internal dosimetry in vivo MDAs by date (continued).

Radiation/

Radionuclide Year

In vivo 

nCi

 In vivo

Count

Time (min) Reference

Am-241 1989-1999 0.6 (lung) 60 Martin 1989, Grothaus 1993, 

Andersen et al. 1995

1993 0.1 (wound) 20 Grothaus 1993

1990-2000 0.2 (lung) Rich 1990

2001-present 0.14 (lung) 60 Rielly 2001

Cm-244 1974-present 1000 ACB 1974

Cf-252 1974-present 1000 ACB 1974

Table 5.5-1. Summary statistics from the 1963 whole-body counting program. 

Radionuclide

Times 

Reported

Number of 

Individuals Maximum Activity (μCi)

Cr-51 15 10 1.2

Co-60 / Fe-59 848 387 1.5

Mn-54 98 51 0.16

Co-58 62 50 0.03

Zn-65 505 171 1.20

Zr-/Nb-95 427 232 1.66

Ru-103-106 93 75 0.22

Ag-110m 583 186 0.93

Sb-122 2 2 0.08

I-131 110 82 5.0

Cs-134 361 168 0.14

Cs-137 2332 573 1.32

Ba-/La-140 90 51 0.07

Ce-141-144 59 49 0.16

Ta-182 50 36 0.02

Hg-203 28 6 0.16

Pa-233 13 10 0.48

Np-239 1 1 1.68

Sb-125 3 3 0.1

Mo-Tc-99 8 5 0.72

I-132 8 7 <0.1

I-133 3 3 <0.1

Te-132 6 6 <0.1

Hg-197 7 3 0.7

5.6 SPECIFIC FACILITIES 

Each of the facilities were responsible for conducting internal dosimetry monitoring programs, which 

were designed both to prevent and/or mitigate internal exposure and to evaluate and document 

internal dose above the detectable limits. The descriptions below provide insight and default 

instructions for dose reconstruction if the specific operational facility is known.

5.6.1

GE built the first facilities at TAN in 1952 for the ANP program, which was active during the 1950s and 

early 1960s before ANP was determined to be impractical. The initial mission was to develop reactors 

for aircraft propulsion. Large facilities built for this program have been used for a number of 

subsequent INEEL projects. Approximately 25 different reactor concepts and experiments have been 

conducted at this location. Large hot cells, maintenance shops, water storage pools, waste 

management areas, etc. are featured (Stacy 2000).

Test Area North
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The Initial Engine Tests (IET) in the 1950s were conducted under area controls and radiological 

monitoring surveillance. The reactors operated in the open environment and each test involved the 

release of large quantities of short-lived radioactive fission product gasses and volatiles. However, 

workers were protected by enclosures (control point buildings, etc.) and constant monitoring for 

identification of unanticipated exposures, etc.

Following termination of the IET programs, the TAN facilities were used to handle, inspect, store, and 

prepare for disposal the materials from unplanned reactor excursions. Major reactor components of 

the damaged SL-1 reactor were examined in the large hot cell and prepared for ultimate disposal. 

Fuel from the damaged Three Mile Island (TMI) reactor was brought to and stored in the large water 

pool facility, where it was examined, and prepared for permanent dry storage. 

The operation of the hot cells, storage basins, and waste treatment facilities involved aged MFPs 

(primarily 137Cs and 90Sr/Y) with periods when 95Zr/Nb and 144Ce were present. Activation products 

were also encountered – primarily 60Co. Alpha emitters (U isotopes and TRU radionuclides) were 

present, but in ratios of at least 50:1 beta:alpha.

Operation of the 15 experimental reactor facilities in this area of the Site resulted in short-lived fission 

products in addition to longer-lived MFPs and some MAPs. A few reactor experiments involved 

operation-to-destruction, with the attendant breach of containment and potential internal exposures.

5.6.1.1 Specific Manufacturing Capability Project

The SMC project began late in the Site history (1985) at the ANP site in the large hangar facility. The 

program uses depleted uranium (DU) to produce armor packages for the U.S. Army M1-A1 and 

M1-A2 main battle tanks (Stacy 2000). During DU parts fabrication, small quantities of finely divided 

U metal and oxides present inhalation and ingestion potential, as indicated by routine positive 

personnel bioassays. 

Air monitoring is the primary method used at the SMC project for evaluating the potential for exposure 

to airborne DU. Fixed-head air sampling throughout the plant, supplemented by CAMs, provides the 

routine information to evaluate the effectiveness of control programs and indicate potential internal 

intake. Exposures to concentrations > 0.1 DAC generally will indicate the use of respiratory 

protection and require bioassay follow-up (King 2001).

The radionuclides of concern at SMC are isotopes of depleted U as listed in Table 5.6.1.-1. The mass 

percentages, relative activities in pCi/μg, and the total pCi/μg are based upon IMBA NIOSH default 

values.

Table 5.6.1.-1. Mass and activity ratios of SMC DU uranium isotopesa 

Isotope Mass % Activity % 

Relative activity 

(pCi μg-1)

Total pCi μg-1

U-238 99.8 83.42 0.3354 0.402

U-236 0.0031 0.05 0.0002

U-235 0.20 1.07 0.004

U-234 0.001 15.46 0.062 a In addition to the U isotopes, two beta-emitting radionuclides are found in DU. Due to short half-lives, 234Th and 234mPa 

reach equilibrium with the 238U parent within about 6 months of billet casting. Beta monitoring is an important part of 

radiological survey and measurements, even though these radionuclides do not contribute significantly to internal doses from 

intakes of DU.
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Inhalation Absorption Type: Respirable particulates associated with SMC operations are probably 

a mixture of metal and metal oxides. The actual exposures are undoubtedly to mixtures of absorption 

types. During the 18 years of operation, much bioassay data has been collected on a large number 

of individuals. The overall elimination patterns are consistent with type M, but probably are a mixture 

of all types. It might be too simplistic to assume a pure absorption type when the chemical form is not 

known for certain. The dose reconstructor should assume either type M or type S to maximize the 

dose to the organ of concern. Exposure to significant quantities of type F uranium at SMC is not 

considered credible.

Particle Size: Detailed particle sizing analyses of representative samples from the various operations 

indicate that an AMAD of 2.4 μm is appropriate for typical SMC operations. This site-specific value of 

2.4 μm AMAD is used for assessments of intakes at the SMC and is the default SMC particle size 

distribution. (Rielly 2001)

Chemical Toxicity: The Threshold Limit Value (TLV) for inhalation of airborne concentrations of 

uranium and its compounds (independent of isotopic composition), as reported by the American 

Conference of Governmental Industrial Hygienists (ACGIH) is 0.2 mg/m3

. The TLV is defined by the 

ACGIH as “...the time-weighted average concentration for a normal 8-hr workday and a 40-hour 

workweek, to which nearly all workers may be repeatedly exposed, day after day, without adverse 

effect.” Converting the mass concentration (using the specific activity of DU of 3.81x10-7 mCi/mg) 

results in a radioactivity concentration of 7.6 x 10-11 μCi/cc. The DAC for Type M U is 3 x 10-10 μCi/cc, 

which is a factor of 4 more than the chemical TLV. The SMC staff has always been aware of the need 

to consider the chemical toxicity of SMC DU exposure in addition to the radiological limit.

Natural Background Uranium Excretion: Urine samples submitted by SMC nonradiation workers in 

1987, 1994, and 1998 were assumed to represent nonoccupational elimination of the SMC worker 

population. The results ranged from 0.04 to 0.33 μg/L with wide fluctuations in individual 

measurement – some as high as 1.0 μg/L (King 2001). The average reported U concentration was 

0.157 ±0.109 μg/L at 1 sigma uncertainty. Therefore, 0.16 μg/L is used as the nonoccupational 

component of U excretion for SMC workers, and is subtracted from each urine result prior to 

assessment of occupational internal dose. The bioassay results in the claimant files reflect the 

subtraction of 0.16 μg/L from the value determined in the laboratory bioassay result.

5.6.2

The ICPP consists of a complex of high-enriched spent fuel storage basins, fuel dissolution and U 

extraction processing facilities, high-level liquid waste storage tank farm, high-level waste calcining 

processes, and associated analytical and support capabilities. The ICPP was a process facility for the 

recovery of high-enriched uranium from spent fuels from a variety of national and a few foreign 

reactors. Because high-enriched uranium was the product, the process vessels had to be small for 

criticality control considerations. Rather than being a plant with large canyon construction and 

complete remote control and maintenance, the ICPP processes were remotely controlled, but contact 

maintenance (i.e., maintenance personnel entered process cells and repaired equipment using 

hands-on maintenance) was required. The process equipment in the cells (5 feet of high-density 

concrete) were decontaminated by flushing/rinsing with concentrated acids and complexing agents 

prior to entry by health physics and maintenance personnel. These occasional operations were well 

planned, but they carried high potential for internal exposures. 

Idaho Chemical Processing Plant (ICPP) – currently INTEC 

The primary internal dose experience at the ICPP resulted from accidental releases. Table 5.6.2-1 

lists unusual and episodic events that have occurred at the ICPP.
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Table 5.6.2-1: Notable Airborne Incidents at ICPP. 

Date Incident

Radionuclide(s)

Released Internal Dose Discussion Reference

05/15/57 Iodine release to Y-Cell I-131

Y-Cell modifications resulted in 8 personnel 

receiving minor thyroid doses in the range of 

600 mrem.

Vance 1957

03/20/58 Iodine release at ICPP I-131

Radioactive iodine spread through makeup 

area to operating corridor. Thyroid intake to 

several HP technicians and operators in the 

40-μCi range.

Rich 1958

Hayden 1958

10/16/59 Criticality accident - in 

shielded process system

Short-lived 

noble gases 

and I-131, -132, 

-133, etc.

Short-lived radioactive gases released to 

plant areas - internal doses reported as 

minimal.

Ginkel et al. 

1960

01/25/61 Criticality accident - in 

shielded process system

Short-lived 

noble gases & I131, -132, -133, 

etc.

Short-lived radioactive gases released 

through process off-gas system to 76-meter 

stack. Internal doses reported as minimal.

Paulus et al. 

1961

01/72 Release of ~1.0 Ci Ru-106 

from ICPP main stack Ru-106 No internal doses detected. ERDA 1977

11/17/72 ICPP mass-spectrometry Pu 

contamination Incident Pu-238, -239

An exposure incident involving about a dozen 

personnel resulted in 50-year exposure lung 

doses ranging up to about 4 rem.

Wenzel 1973

Wenzel 1974

10/17/78 Criticality accident in 

shielded process system

Short-lived 

noble gases

Short-lived lived radioactive gases released 

through process off-gas system to 76-meter 

stack. Internal doses reported as minimal.

Casto 1980

11/85 N-Cell Pu Uptake Pu-238

Internal exposures were far below DOE 

exposure limits, but showed a weakness in 

the radiological control program.

Henry and 

Slagle 1985

10/30/88 Release of ~0.2 Ci of Ru-106 

from ICPP main stack Ru-106 No internal doses detected. Hoff et al. 1989

5.6.2.1 High-Enriched Spent Fuel Storage

The original spent fuel storage facility (CPP-603) was a 1.5-million-gallon, three-basin, 20-foot deep, 

unlined concrete water pit that operated from 1950 until 1984. Because the basin was unlined, the 

use of demineralized water was not feasible, due to corrosion of the bare concrete. In the late 1970s, 

ion exchange columns and a sand filter were installed for water cleanup.

The basin was used for the storage of aluminum-clad, stainless-steel-clad, zirconium-clad, and 

sodium-bonded stainless steel fuels. During 34 years of operation, cases of cladding leaks resulted in 

MFP contamination of the basin water. Breaks in the stainless-steel cladding of sodium-bonded 

elements allowed a sodium-water reaction that increased the dispersal of fission products to the pool 

water. Pool water contamination up to and > 0.5 μCi/mL existed for periods during which cleanup 

methods were developed. As a result, during the late 1950s through the1970s, air activity above the 

pool area was routinely measured in the few tenths to 1 MPCa levels and was one of the very few 

operations at INEEL in which operators were allowed to work in fractional MPCa levels without 

respiratory protection for several hours a day. Routine bioassay sampling was increased for those 

personnel. The primary contaminants were aged MFP – primarily 90Sr/Y and 137Cs. For claimants in 

which it can be established that they worked in the Building 603 storage facility for extended time 

periods, and specific bioassay analyses are either not available or insufficient, a claimant favorable 

default intake of 1000 DAC-hours per year should be assumed, i.e. 1000 hours/year at 2 x 10-9 μCi/cc: 

20,000 cc/min x 60 min/h x 1000 h/y x 2 x 10-9 μCi/cc x 106 pCi/μCi = 2.4 x 106 pCi/y.
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In 1984, the Fluorinel Dissolution Process and Fuel Storage Facility (FAST) with a large stainlesssteel-lined pool with ion exchange cleanup systems, etc., was completed for improved contamination 

control and the reduction of chronic internal exposure potential.

5.6.2.2 High-Level Wastes

The high-level waste storage tank farm consists of a series of 500,000-gallon, underground, stainlesssteel tanks, each of which is in a reinforced concrete bunker with alarmed sumps. Several major 

underground spills occurred in the tank farm, primarily at stainless-mild steel connections. These 

spills resulted in extremely high-levels of contaminated soil that were removed with remote equipment 

and transport. These operations had high release potential, and were planned and executed with 

personnel protective equipment and monitoring. The contaminated soil was kept wet when being 

handled to minimize airborne contamination. Both the in vitro and in vivo data are documented in the 

personnel dosimetry files.

5.6.2.3 High-Level Waste Calcination

High-level waste calcination operations, which began in 1963, consisted of the production of a hightemperature calcine by spraying high-level liquid waste into a fluidized bed of calcine. The nitric 

oxides were vented up the stack following high-efficiency filtering, leaving the radionuclides high-fired 

coated on calcine granules. The potential for release of high-level MFP and MAP and TRU particulate 

activity was significant, but recognized and monitored with an extensive array of air and radiation area 

monitors. Facility operators were placed on routine as well as special bioassay schedules.

5.6.2.4 Process Analytical Facilities

The Remote Analytical Facilities (RAF) and lower level Process Sample Analytical Laboratories 

analyzed samples critical to process controls. The potential for radioactive material release and 

internal exposure was significant in these facilities. Some internal exposure incidents occurred in the 

laboratories from loss of control of process samples in hoods, etc. These incidents frequently 

involved loss of control of TRU materials during analytical procedures in hoods or benchtop 

confinement-type operations.

5.6.2.5 Spent Fuel Processing and ICPP Most Limiting Radionuclides

High-enriched fuel processing, which began in 1953, consisted of nitric acid dissolution of aluminumclad elements, electrolytic nitric acid dissolution of stainless-steel-clad elements, hydrofluoric acid 

dissolution of zirconium-clad elements, and nitric acid dissolution of graphite fuel following graphite 

burning. Uranium was extracted from the dissolved elements (using variations of the PUREX 

process), producing a uranyl nitrate product. The first-, second-, and third-cycle raffinates contained 

high levels of fission and activation products along with TRU. These products in highly corrosive 

matrices were difficult to contain in the confinement barriers of the process piping, process off-gas, 

etc., and resulted in routine leaks and spills. Most of the leaks were confined to process cells, but 

occasionally occurred in occupied process control and equipment areas. For this reason, the 

extensive network of CAMs and Radiation Area Monitors (RAMs) was essential for timely detection of 

loss of confinement. 

Source terms for the various ICPP operations are based on the types of fuel that were reprocessed. 

All of the fuels reprocessed were highly enriched with 235

U enrichments in the expended fuels ranging 

from 50% to 93%. Most of the ICPP reprocessing involved aluminum clad fuels from test reactors, 

stainless steel clad fuels from the Experimental Breeder Reactor II (EBR-II), and zirconium clad fuels 

from various reactors. There were other minor campaigns for processing minor amounts of fuel such 
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-5 Page 32 of 49

as graphite fuel from the ROVER program. In addition, some fuels that had no burnup or very little 

burnup were reprocessed in a hands-on operation known as Custom Processing. 

In order to minimize the radiological safety hazard due to the relatively volatile halogens, the fuel was 

normally decayed a minimum of 120 days before shipment to ICPP for processing. Processing of 

fuels often did not occur until years later. Because of this relatively long decay time, many of the short 

half-life radionuclides experienced considerable decay, leaving the actinides to make up a larger 

percentage of the total radionuclide inventory of the processed fuel. To give an indication of the 

radionuclide inventory in fuels that were processed at the ICPP, the ORIGEN2 code (Croff 1980) was 

used to decay previously calculated fuel inventories (Wenzel 2000) to typical decay times before 

processing. Aluminum fuel, based on fuel from the Advanced Test Reactor (ATR) was decayed 120 

days. Stainless steel fuel, based on fuel from EBR-II, was decayed 3 years. Zirconium fuel, based on 

various reactors, was decayed 5 years. These decay times represent the minimum decay times that 

occurred prior to processing of the fuels. Table 5.6.2.5-1 lists the radiologically significant nuclides 

and the percent of the inhalation dose from inhaling activity from these three types of fuel. The 

inhalation dose percentages were calculated using ICRP 68 5 μm AMAD dose conversion factors. 

Table 5.6.2.5-1. Radiologically significant radionuclides for ICPP processed fuelsa

. 

 

 

Nuclide

 

Half-life

 

Absorption 

Type

Aluminum-Clad Fuel

(decayed 120 d)

Stainless Steel-Clad 

Fuel (decayed 3 y)

Zirconium-Clad Fuel

(decayed 5 y)

Relative 

Activity

Percent

Inhalation 

Dose

Relative 

Activity

Percent

Inhalation 

Dose

Relative 

Activity

Percent

Inhalation 

Dose

Sr-89 50.5 d F 7.8E-2 1.7 1.0E-6 0.0 2.1E-11 0.0

Sr-90 28.78 y F 5.7E-3 2.7 8.6E-2 13.3 2.0E-1 6.0

Y-90 64.00 h S 5.7E-3 0.2 8.6E-2 0.8 2.0E-1 0.3

Y-91 58.51 d S 1.1E-1 10.9 9.7E-6 0.0 8.0E-10 0.0

Zr-95 63.98 d S 1.4E-1 8.9 3.3E-5 0.0 5.6E-9 0.0

Nb-95 35.15 d M/S 2.4E-1 4.9 7.3E-5 0.0 1.3E-8 0.0

Ru-103 39.26 d S 6.10 1.09 0.00 0.00 0.00 0.00

Ru-106 368.2 d S 8.9E-3 4.8 3.1E-2 5.7 4.5E-3 0.2

Cs-134 2.07 y F 9.2E-09 0.7 3.1E-12 0.1 5.0E-09 0.4

Cs-137 30.07 y F 6.3E-3 0.7 9.3E-2 3.2 2.1E-1 1.4

Ce-144 284.3 d M 1.3E-1 46.7 1.6E-1 18.4 2.2E-2 0.5

Pm-147 2.623 y M 1.6E-2 0.9 1.8E-1 3.2 8.4E-2 0.3

U-234 245500 y S 6.3E-10 0.0 1.1E-4 4.0 7.2E-8 0.0

Pu-238 87.71 y M 3.1E-5 14.4 7.6E-6 1.2 3.0E-3 90.0

Pu-239 24110 y M 1.1E-7 0.1 3.0E-4 50.0 1.6E-6 0.1

Totalb 98.6 99.8 99.2

Total of highlighted select nuclides 64.5 86.1 97.9

Weighting factor for the 4 selected 

radionuclides

1.6 1.2 1.0

Mass fraction U-234 to total U 3.4E-5 5.3E-5 1.4E-4 a Assume exposure to aluminum-clad fuel activity from the beginning through1970, zirconium-clad fuel activity 

 from 1971 to present, and stainless steel-clad fuel activity when there is indication that the Pu-239 activity 

 exceeds the Pu-238 activity. b The total percent of the inhalation dose is less than 100% because other radionuclides not included in the table 

contribute small amounts of dose.

Aluminum fuels were processed between 1953 and 1986, stainless steel fuels primarily between 1977 

and 1981, and zirconium fuels between 1972 and 1988 (Staiger 2003). There are relatively long half-
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-5 Page 33 of 49

lived fission products that persist for the ICPP source terms. In most cases, the source terms were 

well tagged with beta emitting radionuclides which allowed beta/gamma-detecting CAMs to be used at 

ICPP with the realization that they would also warn of possible alpha contamination or internal 

exposures. 

Table 5.6.2.5-1 contains too many radionuclides for efficient dose reconstruction. Rather than include 

all of the radionuclides in the default summary table for missed dose (Table 5.7-1), only 90Sr, 137Cs, 

144Ce, and 238Pu are included for aluminum and zirconium fuels. For stainless steel fuels, the 238Pu is 

replaced by 239Pu. Cesium-137 was selected because it is most commonly reported in 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 at the bottom of Table 5.7-1. This gives an equivalent to 100% of the dose from the 

radionuclide distributions for the three types of fuels.

One exception to this planned fuel aging was the RaLa process, which operated in “L” cell of the 

601/602 process building beginning in February 1957 to 1963. This process was designed for the 

ICPP to extract radioactive lanthanum (RaLa) from “green” fuel from the MTR reactor with as little 

decay as manageable (less than 2 days decay). Fuel was removed from the MTR reactor, 

transported about 2 miles to the ICPP in a heavy shielded transport container by a “straddle carrier,” 

immediately dissolved, and the barium element was extracted. The 140Ba/La product was shipped 

immediately to Los Alamos National Laboratory (LANL). This process released large quantities of 

volatile radioactive iodines, which have a higher potential for escaping confinement systems. Several 

significant internal exposure incidents occurred in which 131I, 132I, and 133I thyroid intakes occurred 

before personnel could respond to CAM alarms and take protective or corrective actions.

5.6.3

ANL-W, based at the INEEL since 1951, continues to conduct nuclear programs. Although ANL-W 

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

programs operate under the DOE Chicago field office. However, over the 52 years of ANL-W 

experience, informal program coordination has occurred, such that technical consistency with the 

other INEEL facility internal dosimetry programs exists.

Argonne National Laboratories-West

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

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, U-fueled, Pu-fueled, breeder reactor designs; and selfdestruct experiments. As listed in Table 5.6.3-1, the radionuclides of concern ranged the spectrum of 

MFP, MAP, U, and TRU. 

5.6.4

The RWMC has supported NRTS/INEEL operations as a waste management complex since 1952 

and has received major quantities of TRU waste from Rocky Flats and other DOE facilities. Improved 

operations have resulted in a decrease in internal dose potential. The original disposal techniques, 

dumping waste to open trenches, were relatively vulnerable to airborne release compared to current 

total containment practices. The four major areas in the RWMC facility are the Subsurface Disposal 

Area (SDA) for permanent disposal of low-level waste and some early TRU waste, which will be 

exhumed and repackaged; the 58-acre Transuranic Storage Area (TSA) for temporary storage, 

examination, and certification prior to shipment to the Waste Isolation Pilot Plant (WIPP); operations 

area; and an administrative area, where no radioactive waste is permitted.

Radioactive Waste Management Complex
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-5 Page 34 of 49

Table 5.6.3-1. Radionuclides of concern, ANL-W locations, and MDAs. 

Radionuclides

Absorption 

Type Sources/characteristics

Urine

pCi/mL

In vivoa

5 min count

nCi

In vivob

10 min count

nCi

H-3 (HTO) F EBR-II Reactor Facility & 

Sodium Components 

Maintenance Shop

2.

Mn-54 F, M EBR-II primary source. 

Levels low and decreasing

2.6 1.3

Fe-59 4.5 2.4

Co-58 2.5 1.3

Co-60 2.5 1.4

Sr/Y-90 F All facilities handling fission 

products

0.02c

Cs-134 F All facilities handling fission 

products

2.7 1.3

Cs-137 3.0 1.5

U-235 F, M, S Hot cell, hoods, 

gloveboxes, waste, reactor 

fuel, research areas

0.02c 0.1

U-238 0.02c 1.6

Pu-238 M, S FCF, HFEF, ZPPR, 

Analytical Laboratory

0.02c 60.

Pu-239 0.02c 161.

Am-241 M FCF, HFEF, ZPPR, 

Analytical Laboratory

0.1

a

Using a Canberra WBC with a 5-minute count. b

Using a large NaI WBC Detector with a 10-minute count. c

Assumes a 100 mL sample.

A comprehensive radiation protection program is practiced, which includes extensive air monitoring, 

personnel contamination, and surface contamination surveillance. Though infrequent, there have 

been instances of inadvertent intakes (recently there was one in 1996 and two in 1992). Therefore, 

current routine bioassay is random.

Tables 5.6.4-1 and 5.6.4-2 summarize the major radionuclides in the RWMC waste inventory. TRU 

radionuclides are the primary contaminants in the TSA waste. All but 241Pu are alpha emitters. 

Because the materials are not homogenous, it should not be assumed that no detection of one 

radionuclide in the inventory invalidates detection of other radionuclides. 

Because the waste by nature has all chemical and physical characteristics, the claimant favorable 

absorption type should be assumed unless data indicate that another type fits the data better. 

Exposures in later years will most likely be from contaminated soil or corrosion products and a default 

5 μm AMAD particle size distribution should be assumed as recommended by ICRP 68.

Table 5.6.4-1. Radioactive waste inventory in the TSA. 

Waste type

Volume 

(m3)

Total 

curies Radionuclide

Concentration 

(Ci/m3) Percentage

Stored Contact 

Handled TRU waste

65,000 4.06E+5 Pu-241 2.5E+00 44.1

Am-241 1.4E+00 24.7

Pu-238 9.7E-01 17.1

Pu-239 6.3E-01 11.1

Pu-240 1.5E-01 2.6

U-233 1.4E-02 0.2

Cm-244 0.8E-02 0.1
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-5 Page 35 of 49

Table 5.6.4-2. Radioactive low-level waste inventory in the active pits in the SDA. 

Waste type

Volume 

(m3)

Total 

curies Radionuclide

Concentration 

(Ci/m3) Percentage

LLW 75,600 3.35E+05 Co-60 4.1E+0 92.

Ni-63 3.3E-1 7.4

Sr-90 9.7E-3 0.22

Cs-137 9.7E-3 0.22

H-3 5.8E-3 0.13

C-14 8.9E-4 0.02

5.6.5

The Waste Reduction Operations Complex (WROC) includes several reactor facilities that operated 

from the 1950s to the late 1960s and the Power Burst Facility (PBF), a reactor that operated from 

1972 to 1985. These currently inactive facilities are in a common control area. In addition, a low-level 

waste incinerator called the Waste Experimental Reduction Facility (WERF) burned waste from all 

INEEL facilities from 1982 to 2001. The WERF, which is undergoing decontamination and 

decommissioning (D&D), was a low-level waste incinerator, including some mixed-waste treatment 

(Stacy 2000). 

Waste Reduction Operations Complex

The waste at WERF is in the form of burnable containers and resulting high-fired and solidified ash. 

The radioactive waste at the mixed waste storage facility and the reactors were the sources of the 

radioactivity inventory. The ashes were removed remotely to a glove box and solidified in 55-gallon 

drums. 

The radiological protection program included CAMs and fixed air sampling systems, RAMs, surface 

and personnel contamination surveillance, and effluent monitors.

Since the type of radioactive materials processed at WERF varied, depending on the area shipping 

the waste to WERF, the claimant favorable assumption is that radioactive materials came from Zr fuel 

as processed at the ICPP (see Table 5.6.2.5-1, columns 7 and 8).

5.6.6

5.6.6.1 TRA Reactors

Test Reactor Area

The MTR was the second operating reactor at the NRTS (from March 1953 to 1970); the ETR (1957-

1981) and the ATR (1967-present) along with the six reactor critical facilities supporting the test 

reactors were also at the TRA. The TRA complex includes hot cells, a gamma irradiation pool facility, 

research laboratories, and analytical laboratories. The reactors at TRA, as well as the others at the 

INEEL, were used for testing materials, experiments, neutron irradiation facilities, etc. They were not 

involved in the production of Pu or any other weapons materials as were some DOE reactors. 

The uranium in the TRA reactors is enriched to 93% with 235U. The fuel cladding for the TRA reactors 

is Al. The predominant activation product in the cladding of Al is 24Na, formed by activation of Na 

present in the Al. Sodium-24 has a half-life of 15 hours and emits a high-energy gamma ray (2.75 

MeV). The inhalation dose to personnel from 24Na is insignificant compared to the fission products in 

the fuel. There are minor levels of activation products of stainless steel, 58Co, 60Co, 51Cr, 56Mn, etc., in 

the primary coolant system (PCS) due to corrosion of the stainless-steel PCS components. The 

claimant favorable recommendation is to use any in vivo counting data in the claimant files directly. If 

applicable data is absent, the in vivo MDLs may be assumed, consistent with information in Tables 

5.1.1-1 and 5.4-3. 
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Several factors contributed to unusual amounts of fission products in the PCS of the MTR and ETR 

during early operations. With fuel element cladding technology in its infancy; the quality of the 

cladding was not the best and fission products leaked through the cladding. Another factor was 

“tramp” fuel that was a contaminant on the outside of the cladding. During reactor operation, fission 

products created in tramp U were released directly to the PCS. Reactor operators and the fuel 

manufacturer resolved these deficiencies over time such that by the time the ATR became 

operational, the PCS of the ATR was considerably less contaminated than that of the MTR or the ETR 

during their early years of operation. The radiologically significant radionuclides for Al fuel processed 

at the ICPP in Table 5.6.2.5-1 are also applicable to this PCS contamination.

The majority of radioactivity releases from the TRA reactors to areas potentially occupied by workers 

consisted of noble gases that promptly decayed to short-lived particulate. The principle dose to 

personnel from releases of noble gases was direct radiation, not inhalation. The direct radiation 

caused instruments to alarm, resulting in immediate evacuation of the affected areas. The claimant 

favorable position is that there could have been some halogens and particulate radionuclides released 

along with the noble gases. The ORIGEN-2 code (Croff 1980) was used to calculate a radionuclide 

inventory for ATR fuel at the time of reactor shutdown. The radionuclide inventory was then 

fractionated according to the release percentages often used to assess hypothetical reactor accidents 

(AEC 1962b): 100% noble gases, 50% halogens and 1% solids. A halogen to solid ratio of 50:1 was 

therefore used to determine the radionuclides that could significantly contribute to an inhalation dose. 

Table 5.6.6.1-1 contains the relative amounts of halogens and particulate radionuclides from ATR fuel 

that contribute significantly to the inhalation dose. Table 5.6.6.1-1 contains too many radionuclides for 

efficient dose reconstruction. As shown in Table 5.6.6.1-1, the iodine isotopes contribute by far the 

majority of the dose from a test reactor gaseous release, 94.5 percent. Therefore, I-131 has been 

selected to be the representative radionuclide for missed dose in Table 5.7-1, and the dose from 131I is 

weighted by a factor of 1.6 to account for the total iodine dose. When there is measured I-131 for any 

facility, a dose reconstructor may choose to apply this factor of 1.6 to the I-131 dose to account for 

shorter lived iodines and may choose to ratio Ce-144, Sr-89, Y-91, and Zr-95 activities to the I-131 

activity using the values in Table 5.6.6.1-1 

Table 5.6.6.1-1. Radiologically significant radionuclides for ATR fuel 

gaseous releases. 

Nuclide Half- life Absorption 

Type

Fraction

Activity

Ratio 

Activity to 

I-131 

Activity

Sr-89 50.5 d S 0.0028 3.5E-02

Y-91 58.51 d M/S 0.0033 4.1E-02

Zr-95 63.98 d S 0.0036 4.4E-02

I-131 8.041 d F 0.0816

I-132 2.3 h F 0.1298

I-133 20.8 h F 0.2040

I-134 0.876 h F 0.2225

I-135 6.61 h F 0.1855

Ce-144 284.3 d S 0.0012 1.5E-02

 Total 0.834

Another potential source of inhalation dose was from the pressurized loop experiments in the TRA 

reactors. These loop experiments sometimes contained Ag and Ta, which became activated, 

producing 110mAg and 182Ta. When a depressurizing incident occurred in the loop, these activation 
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-5 Page 37 of 49

products were sometimes released to work areas. One of the major airborne incidents at TRA listed 

in Table 5.6.6.1-2 involved the release of 182Ta. The claimant favorable position is to assign a missed 

dose based on the MDA for in vivo counting (see Table 5.4-3) for 110mAg and 182Ta.

Reactor components, such as fuel elements, reactor loop components, etc., that are removed from 

the reactor and placed in the canal are a source of contamination in the canal water. If these 

components are not cleaned adequately before they are removed from the canal, the activity on them 

can become airborne. The radiologically significant radionuclides for listed in Table 5.6.2.5-1 for Al 

fuel processed at the ICPP, are applicable to this TRA contamination also.

Table 5.6.6.1-2. Major airborne incidents at TRA. 

Date Incident

Radionuclide(s) 

released Reference

3/28/1954 GE-ANP-1 depressurization Noble gas Sommers, 1954

12/17/1958 GEH-4 rupture Noble gas + I Sommers 1958

6/13/1967 GA-18-1 depressurization Ta-182 and Ta-183 Nertney et al. 1967

1/6/1967 Noble gas release at ATR Noble gas Sommers 1977

5.6.6.2 TRA Laboratories

The wing buildings of the MTR at TRA house chemistry and other laboratories. Over the years, 

experiments in the laboratories resulted in contamination and airborne activity incidents. These 

experiments involved various isotopes of Pu, U, and other radionuclides. The radiologically significant 

radionuclides listed in Table 5.6.2.5-1 for Al fuel processed at the ICPP are applicable for the 

experiments done in the TRA laboratories.

Beginning in 1980, many of the laboratories were involved in various assays of TMI reactor fuel from 

the accident in March 1979. Because the TMI core had only the equivalent of 100 effective full-power 

days at the time of the accident, the TRU content would not be as high as that for a typical power 

reactor core, which is routinely operated for a year or so before the fuel is changed. Use of the 

radiologically significant radionuclides for Zr fuel processed at the ICPP in Table 5.6.2.5-1 is claimant 

favorable for evaluating inhalation dose from exposure to contamination from the TMI fuel.

5.7 DEFAULT FOR MISSED DOSE

Based upon the INEEL characteristics and circumstances, a number of missed dose default 

assumptions have been derived. Table 5.7-1 is a summary of these recommended defaults. The 

bases for these defaults are discussed in more detail in Table 5.7-1 below: 

For most of the history of the INEEL personnel dosimeters were issued to all workers who entered the 

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

administrative and clerical personnel were required to wear these radiation monitoring dosimeters 

though they were not exposed to elevated backgrounds or internal dose potential. The default 

categories listed in the Table 5.7-1 are determined by the following: If claimant file includes positive 

external dosimeter readings, they should be treated as radiation workers and the default internal 

missed dose is applied as outlined in the table. If no detectable external or internal dose information 

is recorded, only the environmental dose should be included.
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-5 Page 38 of 49

Table 5.7-1. Default table for missed dose

Time Period Based On Recommendation Site Area Basis

Start up date 

through 

1960

Urine

gross β

Calculate chronic Sr-90 intake that 

results in a urine activity of 0.4 x 

gross β

All 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 Cs137 & Sr-90 are approximately 

equal.

Pu-238 intake = 0.008 x Sr-90 

intake

Pu:Sr-90 ratio of 0.005 in Al fuel 

weighted by a factor of 1.6. See 

Table 5.6.2.5-1, column 5.

Ce-144 intake = 35 x Sr-90 intake Ce-144:Sr-90 ratio of 22 in Al fuel 

weighted by a factor of 1.6. See 

Table 5.6.2.5-1. 

I-131 = 5 μCi in vivo TRAa Assumes I-131 in vivo is equal to 

the maximum in vivo measured in 

1963. 

1961 – 1970 in vivo

Cs-137

Calculate chronic Cs-137 intake 

that results in the in vivo

measurement

All

Sr-90 intake = Cs-137 intake when 

no in vitro measurement

Half-lives and fission yields are 

approximately equal.

Pu-238 intake = 0.008 x Cs-137 

intake when no in vitro

measurement

Pu:Cs-137 ratio of 0.005 in Al fuel 

weighted by a factor of 1.6. See 

Table 5.6.2.5-1, column 5.

Ce-144 intake = 35 x Cs-137 intake 

when no measurement

TRA Ce-144:Sr-90 ratio of 22 in Al fuel 

weighted by a factor of 1.6. See 

Table 5.6.2.5-1, column 5.

I-131 = 19 nCi in vivo when no 

measurement. 

MDA for I-131 of 12 nCi weighted 

by a factor of 1.6. See Table 5.4-3. 

Acute intake of Ag-110m = 12 nCi 

in vivo when no measurement. 

Assume 3 days post intake. 

MDA for Ag-110m, see Table 5.4-

3. Known incident involving Ag110m.

Acute intake of Ta-182 = 12 nCi in 

vivo when no measurement. 

Assume 3 days post intake. 

MDA for Ta-182, see Table 5.4-3.

Known incident involving Ta-182

1971-1980 in vivo 

Cs-137

Sr-90 intake = 

Cs-137 intake when no in vitro

measurement

All Half-lives and fission yields of Cs137 & Sr-90 are approximately 

equal.

Pu-238 intake = 0.02 x Cs-137 

intake when no in vitro

measurement

Pu:Sr-90 ratio of 0.02 in Zr fuel. 

See Table 

5.6.2.5-1, column 9.

Ce-144 intake = 35 x Cs-137 intake 

when no measurement

INTECb & 

unknownc 

Ce-144:Sr-90 ratio of 22 in Al fuel 

weighted by a factor of 1.6. See 

Table 5.6.2.5-1, column 5.

Pu-239 intake = 0.004 x Cs-137 

intake when no in vitro

measurement

ANL Pu:Cs-137 ratio of 0.003 in 

stainless steel fuel weighted by a 

factor of 1.2. See Table 5.6.2.5-1, 

column 7 a The dose recommendations specific areas, i.e., TRA, INTEC, ANL, are to be added to recommendations for “All”. b Formerly known as ICPP. c When it is not known what site area a worker worked at, or if they worked at many different areas, use the

 dose recommendations for INTEC.
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-5 Page 39 of 49

Table 5.7-1. Default table for missed dose (continued)

Time Period Based On Recommendation Site Areas Basis

1971-1980 

(continued)

in vivo 

Cs-137 

(continued)

Ce-144 intake = 2.4 x Cs-137 

assigned intake when no

measurement

ANL 

(continued)

Ce-144:Cs-137 ratio of 2 in 

stainless steel fuel weighted by a 

factor of 1.2. See Table 5.6.2.5-1, 

column 5.

Pu-238 intake = 0.008 x assigned 

Cs-137 intake when no in vitro

measurement

other areaa Pu:Cs-137 ratio of 0.005 in Al fuel 

weighted by a factor of 1.6. See 

Table 5.6.2.5-1, column 5.

Ce-144 intake = 35 x Cs-137 

assigned intake when no

measurement

Ce-144:Sr-90 ratio of 22 in Al fuel 

weighted by a factor of 1.6. See 

Table 5.6.2.5-1, column 5.

I-131 = 19 nCi when not reported TRA MDA for I-131 of 12 nCi weighted 

by a factor of 1.6. See Table 5.4-3. 

1981 -

present

Bioassay Calculate associated radionuclide 

intakes based on Al fuel 

distribution

all except 

ANL, 

INTEC & 

SMC

Distribution calculated for Al fuel. 

See Table 5.6.2.5-1, column 3.

Calculate associated radionuclide 

intakes based on stainless steel 

fuel distribution

ANL Distribution calculated for stainless 

steel fuel. See Table 5.6.2.5-1, 

column 4.

Calculate associated radionuclide 

intakes based on Zr fuel 

distribution

INTEC & 

unknown

Distribution calculated for Zr fuel. 

See Table 5.6.2.5-1, column 5.

Calculate associated intakes based 

on known isotope distribution

SMC Isotope distribution, see Table 

5.6.1-1.

a “Other area” or areas are known, just not one of the specific areas with special recommendations. 

5.8 UNMONITORED WORKERS

As noted above, INEEL personnel dosimeters were issued to all workers at facilities handling 

radioactive material. Many of these workers, due to the nature of their work, would not have been 

exposed to internal activity and would not have been subjected to routine bioassay. The INEEL has 

had a rigorous radiation protection program to detect the presence of and the spread of radioactive 

contamination. Areas with potential airborne contamination were monitored with alarming continuous 

air monitors. The spread of contamination was monitored through the use of smears and monitoring 

instrumentation. It was the policy to screen radiation workers through urine samples and whole body 

counting for internal contamination many (often 4 or more) times a year. Most of the activity 

encountered was well tagged with β/γ activity which would have produced measurable direct radiation 

doses on a personnel dosimeter. Due to the rigor of the radiation monitoring program, the probability 

that a worker receiving a significant unmonitored internal intake of radioactive material is very low. It 

is recommended that workers who have no recorded internal dose and wore a personnel dosimeter 

be treated the same as a worker who was monitored but had no bioassay results exceeding reporting 

levels. It is further recommended that individuals that were not issued a personal dosimeter and have 

no record of internal dose monitoring be assigned only the environmental dose for the facility.

Construction workers often worked at the INEEL. 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 the same as a radiation worker. Construction workers that were issued personnel 

dosimeters should be treated the same as facility employees that were issued personnel dosimeters. 

Construction workers that were not issued a personnel dosimeter should be assigned the 

environmental dose for the facility.
Effective Date: 08/11/2004 Revision No. 00 Document No. ORAUT-TKBS-0007-5 Page 40 of 49

REFERENCES

ACB (Analytical Chemistry Branch) 1974, Detection Limits, Health Services Laboratory, December.

ACC (American Cyanamid Company) 1952, CPP Health Physics Manual, IDO 14055, Atomic Energy 

Division, Idaho Falls, Idaho.

AEC (U.S. Atomic Energy Commission), 1958, Permissible Levels of Radiation Exposure, AEC 

Manual Chapter 0524, January 9, 1958; effective February 1, 1958.

AEC (U.S. Atomic Energy Commission), 1959, Annual Report of Health & Safety Division, 1958, IDO12012, Idaho Operations Office, Idaho Falls, Idaho.

AEC (U.S. Atomic Energy Commission), 1960, Annual Report of Health & Safety Division, 1959, IDO12014, Idaho Operations Office, Idaho Falls, Idaho.

AEC (U.S. Atomic Energy Commission), 1961, Annual Report of Health & Safety Division, 1960, IDO12019, Idaho Operations Office, Idaho Falls, Idaho.

AEC (U.S. Atomic Energy Commission), 1962a, Annual Report of Health & Safety Division, 1961, 

IDO-12021, Idaho Operations Office, Idaho Falls, Idaho.

AEC (U.S. Atomic Energy Commission), 1962b, Calculation of Distance Factors for Power and Test 

Reactor Sites, TID-14844, March, 1962.

AEC (U.S. Atomic Energy Commission), 1963a, Annual Progress Report 1962, Health & Safety 

Division, IDO-12033, Idaho Operations Office, Idaho Falls, Idaho.

AEC (U.S. Atomic Energy Commission), 1963b, Standards for Radiation Protection, AEC Manual 

Chapter 0524, Approved August 12, 1963.

AEC (U.S. Atomic Energy Commission), 1964, Health & Safety Division Annual Progress Report 

1963, IDO-12037, Idaho Operations Office, Idaho Falls, Idaho.

AEC (U.S. Atomic Energy Commission), 1966, Some Guidelines for Studies Involving Internal 

Administration of Radioactive Materials to Human Volunteers, IDO-12058, C. W. Sill, Health 

and Safety Division, Idaho Operations Office, Idaho Falls, Idaho, October. 

AEC (U.S. Atomic Energy Commission), 1968, Standards for Radiation Protection, AEC Manual 

Chapter 0524 and Appendix, Approved November 8, 1968.

AEC (U.S. Atomic Energy Commission), 1969, Annual Report 1968, Health Services Laboratory, IDO12071, Idaho Operations Office, Idaho Falls, Idaho.

AEC (U.S. Atomic Energy Commission), 1971, Annual Report 1970, Health Services Laboratory, IDO12075, Idaho Operations Office, Idaho Falls, Idaho.

AEC (U.S. Atomic Energy Commission), 1972, Annual Report 1971, Health Services Laboratory, IDO12076, Idaho Operations Office, Idaho Falls, Idaho.

AEC (U.S. Atomic Energy Commission), 1975, Occupational Radiation Exposure Information, AEC 

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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.

calcine

The dry solid (grainy or granular) product of a chemical process of removing liquids 

from a solution; also, the process for creating the chemical reaction that removes 

liquids from a solution.

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 doesn’t belong. 

core

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

reactor operation.

criticality

An event which occurs as a result of uncontrolled assembly of fissile material capable 

of sustaining a nuclear chain reaction.

curie

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

decontaminate

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

Uranium that has undergone a process to remove one or more isotopes from the 

element. Depleted uranium is usually depleted in 235U and 234U, leaving primarily 238U .

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. 

enriched uranium

Uranium processed to increase the relative abundance of the isotope 235U. Most 

processing methods increase the relative abundance of 234U, also.

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.

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 keV to 9 MeV.

half-life

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

(disintegrate).

hot cell

A specialized shielded laboratory in which radioactive materials may be handled with 

the aid of remotely operated manipulators. The walls and windows of the laboratory 

are made of materials designed to protect workers from radiation. 

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 
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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.

nucleus 

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

mass. 

photon

A quantum of electromagnetic energy often referred to as X-rays or gamma rays, but 

also including light and radiant heat. 

proton

An elementary atomic particle with a positive electrical charge equal numerically to the 

charge of the electron and a mass slightly greater than 1 mass unit.

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). 
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RaLa

Radioactive lanthanum, one of the fission products of a nuclear reaction; a lanthanum 

recovery process at the INTEC for development weapons.

rem

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

factor.

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.

spent fuel storage basin

A pool or pit made of reinforced concrete containing water and used to store spent 

nuclear fuel. The water acts as shielding and as a coolant 

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 years 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.