ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 67

DOE Release Date 02/01/2010

Document Title:

Idaho National Laboratory and Argonne National 

Laboratory - West – Occupational Environmental 

Dose

Document Number: ORAUT-TKBS-0007-4

Revision: 02

Effective Date: 01/08/2010

Type of Document: TBD

Supersedes: Revision 01, ORAUT-

 TKBS-0026-4 Rev 01

Subject Expert(s): N/A

Site Expert(s): Henry Peterson and Norman D. Rohrig

Approval: Signature on File Approval Date: 12/22/2009

Jo Ann M. Jenkins, Document Owner

Concurrence: Signature on File Concurrence Date: 12/22/2009

John M. Byrne, Objective 1 Manager

Concurrence: Keith A. McCartney Signature on File for Concurrence Date: 12/31/2009

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 01/04/2010

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 01/08/2010

Stuart L. Hinnefeld, Interim Director OCAS

New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 2 of 67

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION

NUMBER DESCRIPTION

03/30/2004 00 New Technical Basis Document for the Idaho National Engineering 

and Environmental Laboratory – Occupational Environmental Dose. 

First approved issue. Initiated by Norman D. Rohrig.

08/17/2007 01 Approved Revision 01 initiated to add additional information on how 

TLD results were chosen. Added Purpose, Scope, and Acronyms 

sections. Added references and SRDB numbers. Changed Idaho 

National Engineering and Environmental Laboratory to Idaho 

National Laboratory. Incorporates formal internal and NIOSH review 

comments. Adds Attributions and Annotations section. Incorporates 

formal internal and NIOSH review of the Attributions and Annotations 

section. Constitutes a total rewrite of document. Training Required: 

As determined by the Task Manager. Initiated by Jo Ann M. Jenkins.

01/08/2010 02 Revised to combine ANL-W TBD into the INL TBD. Updated Intake 

and External Dose tables to add data from 2006 to 2008. Added 129I 

intake values to intake tables for all areas. Incorporates formal 

internal review comments. No changes were needed as a result of 

formal NIOSH review. Constitutes a total rewrite of the document. 

Training required: As determined by the Objective Manager. 

Initiated by Jo Ann Jenkins.
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 3 of 67

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

4.1 Introduction ..............................................................................................................................8

4.1.1 Purpose........................................................................................................................9

4.1.2 Scope...........................................................................................................................9

4.2 Internal Intakes from Onsite Airborne Radionuclide Concentration......................................... 13

4.2.1 Operational Releases ................................................................................................. 13

4.2.2 Episodic Releases at INL............................................................................................ 28

4.2.2.1 Stationary Low-Power Reactor No. 1 Accident............................................. 28

4.2.2.2 Criticalities.................................................................................................... 29

4.2.2.2.1 INTEC Criticality of October 16, 1959 ......................................... 30

4.2.2.2.2 INTEC Criticality of January 25, 1961 ......................................... 32

4.2.2.2.3 INTEC Criticality of October 17, 1978 ......................................... 32

4.2.2.2.4 HTRE No. 3 Criticality Excursion (IET 13)................................... 34

4.2.2.3 Releases from Planned Tests ...................................................................... 34

4.2.2.3.1 Fuel Element Burn Test A........................................................... 35

4.2.2.3.2 Fuel Element Burn Test B........................................................... 35

4.2.2.3.3 FECF Filter Break ....................................................................... 35

4.2.2.3.4 Initial Engine Test 14 .................................................................. 35

4.2.2.3.5 Initial Engine Test 15(B).............................................................. 37

4.2.2.3.6 Initial Engine Test 16 .................................................................. 37

4.2.2.3.7 Initial Engine Test 17(B).............................................................. 39

4.2.2.3.8 Initial Engine Test 18 .................................................................. 39

4.2.2.3.9 Initial Engine Test 19(A).............................................................. 40

4.2.2.3.10 Initial Engine Test 25(A).............................................................. 40

4.2.2.3.11 Initial Engine Test 25(B).............................................................. 41

4.2.2.3.12 Initial Engine Test 26(A).............................................................. 41

4.3 External Dose ........................................................................................................................41

4.3.1 Facility Fenceline Annual Doses................................................................................. 42

4.3.2 Facility Air Immersion Doses ...................................................................................... 45

4.4 Uncertainty.............................................................................................................................45

4.5 Attributions and Annotations ..................................................................................................46

References ........................................................................................................................................50

Glossary ............................................................................................................................................61
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LIST OF TABLES

TABLE TITLE

4-1 Intake by year for ANL-W, 1952 to 2008 ................................................................................

PAGE

16

4-2 Intake by year for ARA, 1952 to 2008..................................................................................... 17

4-3 Intake by year for CFA, 1952 to 2008..................................................................................... 19

4-4 Intake by year for INTEC, 1952 to 2008 ................................................................................. 20

4-5 Intake by year for RWMC, 1952 to 2008 ................................................................................ 22

4-6 Intake by year for SPERT, 1952 to 2008................................................................................ 23

4-7 Intake by year for TAN, 1952 to 2008..................................................................................... 25

4-8 Intake by year for TRA, 1952 to 2008..................................................................................... 26

4-9 Comparison of calculated facility intakes with intakes from environmental 

monitoring results................................................................................................................... 29

4-10 Intakes for criticalities that occurred at INL............................................................................. 33

4-11 Intakes for special tests at INL ............................................................................................... 36

4-12 Intakes for initial engine tests at INL....................................................................................... 38

4-13 INL facility fence direct gamma values ................................................................................... 43

LIST OF FIGURES

FIGURE TITLE

4-1 INL site map...........................................................................................................................

PAGE

10

4-2 Chronology of facilities and programs at the INL site.............................................................. 12

4-3 Nine-year (1974 to 1983 excluding 1978) average mesoscale dispersion 

isopleths of air concentration at ground level, normalized to unit release rate ........................ 14

4-4 Onsite and distant particulate beta concentrations in air ........................................................ 30

4-5 CFA air-monitoring data for 1976 ............................................................................................ 31

4-6 Dispersion coefficient contours for the SL-1 accident............................................................. 32

4-7 Example of onsite TLD monitoring data.................................................................................. 44
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 5 of 67

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AMAD activity median aerodynamic diameter

ANL-W Argonne National Laboratory–West

ARA Army (later Auxiliary) Reactor Area

BORAX Boiling-Water Reactor Experiment

Bq becquerel

CERT Controlled Environmental Radioiodine (later Release) Test

CFA Central Facilities Area

CFR Code of Federal Regulations

Ci curie

CPP Chemical Processing Plant 

d day

DOE U.S. Department of Energy

EBR Experimental Breeder Reactor

ECF Expended Core Facility

EEOICPA Energy Employee Occupational Illness Compensation Program Act of 2000

EFS Experimental Field Station

EMR Environmental Monitoring Report

EOCR Experimental Organic-Cooled Reactor

F fast (solubility rate)

fCi femtocurie

FEBT Fuel Element Burn Test

FECF Fuel Element Cutting Facility

FPFRT Fission Product Field Release Test

ft foot

GE-ANP General Electric–Aircraft Nuclear Propulsion (program)

GCRE Gas-Cooled Reactor Experiment

GSD geometric standard deviation

H&S Health and Safety (division)

hr hour

HTRE Heat Transfer Reactor Experiment

ICPP Idaho Chemical Processing Plant

ICRP International Commission on Radiological Protection

IDO Idaho Operations Office

IET Initial Engine Test

in. inch

INEEL Idaho National Engineering and Environmental Laboratory

INEL Idaho National Engineering Laboratory

INL Idaho National Laboratory, including ANL-W

INTEC Idaho Nuclear Technology and Engineering Center

km kilometer

kW kilowatt
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LOFT Loss-of-Fluid Test

LPTF Low Power Test Facility

m meter

M moderate (solubility rate)

mi mile

mL milliliter

mo month

mR milliroentgen

mrem millirem

MW megawatt

NRF Naval Reactors Facility

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NOAA National Oceanic and Atmospheric Administration

NRTS National Reactor Testing Station

NSA North Storage Area

PBF Power Burst Facility

pCi picocurie

POC probability of causation

RSAC Radiological Safety Analysis Computer (program)

RWMC Radioactive Waste Management Complex

S slow (solubility rate)

s second

SL-1 Stationary Low-Power Reactor No. 1

SPERT Special Power Excursion Reactor Test

SRDB Ref ID Site Research Database Reference Identification

STPF Shield Test Pool Facility

TAN Test Area North

TBD technical basis document

TLD thermoluminescent dosimeter

TRA Test Reactor Area

TREAT Transient Reactor and Experiment Test (facility)

TSF Technical Support Facility

U.S.C. United States Code

WERF Waste Experimental Production Facility

WWP Warm Waste Pond

yr year

ZPPR Zero Power Plutonium (later Physics) Reactor

α alpha activity

β beta

γ gamma
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 7 of 67

μCi microcurie

°F degrees Fahrenheit

§ section or sections
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 8 of 67

4.1 INTRODUCTION

Technical basis documents and site profile documents are not official determinations made by the 

National Institute for Occupational Safety and Health (NIOSH) but are rather general working 

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used 

to assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

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

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

employer facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees 

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

EEOICPA defines a DOE facility as “any building, structure, or premise, including the grounds upon 

which such building, structure, or premise is located ... in which operations are, or have been, 

conducted by, or on behalf of, the Department of Energy (except for buildings, structures, premises, 

grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 

7384l(12)]. Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted 

above, any facility that performs or performed DOE operations of any nature whatsoever is a DOE 

facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as 

determined in accordance with the POC [probability of causation1

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

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

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

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section 

of EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42 

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

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

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

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

external occupational radiation exposures are considered valid for inclusion in a dose reconstruction. 

No efforts are made to determine the eligibility of any fraction of total measured exposure for inclusion 

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

occupationally derived (NIOSH 2007):

] guidelines established under 

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

guidelines (nor the dose reconstruction regulation, 42 C.F.R. Pt. 82) define “performance of duty” for 

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2007).

• Background radiation, including radiation from naturally occurring radon present in 

conventional structures

 1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 9 of 67

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic 

reasons

4.1.1

This TBD provides the basis for the environmental doses at the Idaho National Laboratory (INL) and 

its predecessor organizations [the National Reactor Testing Station (NRTS), the Idaho National 

Engineering Laboratory (INEL), and the Idaho National Engineering and Environmental Laboratory 

(INEEL)] and at the Argonne National Laboratory – West (ANL-W). This information can be used in 

dose reconstructions for the EEOICPA to supplement information in an individual’s dose record. 

Purpose

This TBD addresses radioactive material releases from areas or facilities at the INL that could affect 

employees at another INL facility and the ANL-W. The releases discussed here have been divided 

into (1) normal “chronic” operational releases and (2) episodic releases that generally have been of 

short duration (DOE 1991). These radioactive material releases potentially represent unrecorded or 

missed doses, either as direct gamma or beta-gamma from immersion in the radioactive gaseous 

cloud for individuals who did not have personal dosimetry to record the dose or as internal doses from 

inhalation of the materials.

This TBD also addresses direct gamma doses from facility operations. In general, these doses, if not 

controlled by management, increase with time and create a facility background dose. At INL, facility 

background doses were recorded by film badges infrequently and inconsistently before 1970 and by 

thermoluminescent dosimeters (TLDs) on a routine basis since 1972. These doses, or facility 

fenceline doses, as they are sometimes called, are a nebulous indication of a dose that individuals

could receive if they worked at the facility. INL facility fenceline doses (minus background) are 

presented for 11 locations. This TBD presents ANL-W facility fenceline doses (minus background) for 

the Environmental Breeder Reactor No. I (EBR-I) location for 1952 to 1972 and the Transient Reactor 

Test (TREAT) facility and EBR-II locations, for 1972 to the present. 

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 4.5.

4.1.2

The INL site was chosen by the U.S. Atomic Energy Commission (AEC) as an isolated location for 

testing various reactor concepts. INL is isolated from the public in two important aspects: (1) it is 

remotely located from population centers, and (2) it is hydrologically isolated because no surface 

streams originate on and flow to an offsite location and no streams cross the Site. Although INL sits 

above the large Snake River Aquifer that eventually surfaces and enters the Snake River in the 

Hagerman Valley area, the annual flow rate of the water in the aquifer is measured at 5 to 15 ft/d

(ERDA 1977a).

Scope

ANL-W is a unique facility at INL. Although inside the INL boundary, ANL-W is under the jurisdiction 

of the DOE Chicago Operations Office while INL is under the jurisdiction of the DOE Idaho Operations 

Office. Although the facility operates in accordance with 10 CFR. Part 835, its operations and 

atmosphere have been and are more in line with a university engaged with pure nuclear energy 

research; these operations support those of the University of Chicago and Argonne National 

Laboratory–East, near Chicago, Illinois. During its first 14 years of operation (i.e., from 1951 through 

1965), ANL-W was in the southwest corner of INL at the location of EBR-I. At this location, ANL-W 

conducted tests on EBR-I, Zero Power Physics (previously plutonium) Reactor (ZPPR) No. 3, 

Argonne Fast Source Reactor, and all Boiling Water Reactor Experiments (BORAX). In 1958, 

construction began on the TREAT Facility and on EBR-II at the present ANL-W location at the 
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 10 of 67

southeast corner of INL. Since the mid-1970s, essentially all ANL-W operations have been conducted 

at the present location, as shown in Figure 4-1.

Figure 4-1. INL site map (Hoff, Chew, and Rope 1987, p. 46).

During the 50-year history of the site, about 50 different reactor concepts have been designed, built, 

and operated at INL. All of these reactors have been prototype, low-power critical, or test reactors; no 

weapons production or commercial power reactors have been operated at INL. Most, if not all, of 

these reactors have used highly enriched (93% or higher) uranium as fuel. Only a few have produced 

significant airborne effluent: (1) the Heat Transfer Reactor Experiment (HTRE) reactors, operated 

under the General Electric–Aircraft Nuclear Propulsion (GE-ANP) Program at the north end of the site 

at Test Area North (TAN), (2) test reactors (Materials Test Reactor, Engineering Test Reactor, and 

Advanced Test Reactor), all at the Test Reactor Area (TRA) near the middle southern end of the site, 

SOUTH 

SECURITY GATE

▪

▪ ▪ CERRO GRANDE

FRENCHMAN’S

 CABIN

▪

RENO RANCH
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 11 of 67

and (3) EBR-II, at ANL-W at the southeast corner of the site (Osloond 1965; Osloond and Newcomb 

1969).

Another historically important airborne effluent producer is the Idaho Nuclear Technology and 

Engineering Center (INTEC), formerly known as the Idaho Chemical Processing Plant (ICPP). This 

facility, constructed in the early 1950s, began processing nuclear fuel in February 1953 and continued 

until 1992 (Knecht et al. 1997, p 4). Throughout its history, the Chem Plant, or CPP as it is commonly 

known, has reprocessed fuel from test reactors at INL, zirconium-clad fuel reclaimed from various 

reactors, stainless-steel-clad fuel from EBR-II, and many AEC test reactors from around the world. 

Apart from the GE-ANP Program, which is discussed below, INTEC airborne releases have 

historically been the most radiologically significant releases at INL (DOE 1991; Till et al. 2002). 

Through the years that INL Environmental Monitoring Reports (EMRs) have been published, INTEC

airborne effluents have been attributed to creating the maximum INL boundary dose. Considering this 

fact, it should be suspected that INTEC airborne effluent would also be responsible for the maximum 

INL worker doses. Calculations performed for this TBD show that although ICPP airborne effluent is 

the most radiologically significant release at INL, the impact to all facility workers is significantly below 

the allowable and acceptable limits (Peterson 2004). Figure 4-1 shows the locations of INL facilities.

From the beginning of operations at the INL site, facility locations were chosen to limit the potential for 

operational releases at one facility to affect another facility. Because the site encompasses 890 mi2

, 

there was ample room to place facilities with this principle in mind. Because the site has an average 

elevation of 5,000 ft and is generally meteorologically characterized with a nocturnal inversion from 

the north-northeast and a daytime lapse condition with winds from the southwest, transitional weather 

regimes are less frequent than at lower elevations. The 50-year history of the site has demonstrated 

that the large expanse of INL and this meteorological characteristic have been satisfactorily effective 

in maintaining the operational isolation of each facility [1].

Beginning with the GE-ANP Program, which began in the early 1950s, the site has had the capability 

of plume tracking by aircraft. The local National Oceanic and Atmospheric Administration (NOAA) 

field office, which was dedicated to site needs and requirements, provided plume projection 

capabilities with a rather extensive network of meteorological monitoring stations (Yanskey, Markee, 

and Richter 1966). The plumes from all intentional planned releases [Controlled Environmental 

Radioiodine (later Release) Test (CERT), Fission Product Field Release Test (FPFRT), Fuel Element 

Burn Tests (FEBTs) A and B, etc.] were directed over an instrumented monitoring grid (GRID III), 

remotely located from other facilities, such that other onsite facilities were not affected by the release

[2]. In general, these tests were performed in support of a specific program (i.e., the FPFRT and 

FEBTs A and B were conducted to support the GE-ANP Program).

All of the airborne releases at INL that have occurred since the beginning of operations were reviewed 

and analyzed as a result of a request from the DOE Idaho Operations Office (IDO). This request was 

to evaluate the radiological impact to INL boundary individuals from airborne releases that had 

occurred since the beginning of operations at the site. With the help of NOAA, which had hourly 

meteorological data from 1956 to 1988, analyses were completed for all airborne releases that 

occurred at INL. The radiological consequences for an adult, a child, and an infant were calculated 

with Version 4 of the Radiological Safety Analysis Computer (RSAC-4) program (Wenzel 1990). The 

results of the study were published in Idaho National Engineering Laboratory Historical Dose 

Evaluation (DOE 1991). All releases considered for that report are the bases for the releases 

considered in this TBD. In addition, all releases documented in the evaluation, operational and 

episodic, have been independently reviewed and found, with minor modifications, to be substantially 

appropriate. The review, conducted by the Radiological Assessment Corporation at the request of the 

Centers for Disease Control and Prevention and the State of Idaho, also evaluated the methodology 

by which the RSAC-4 program performs dose calculations against the National Council on 

Radiological Protection and Measurements (NCRP) methodology (Till et al. 2002). It stated: 
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 12 of 67

As a final point, Tables 7, 8, 9a, 9b, 10a, and 10b, and Figures 18 and 19 confirm that 

the NCRP method was suitable for these ranking purposes when the results are 

compared with those using the RSAC code. In all cases, the RSAC code confirmed 

the results obtained using the NCRP methodology. (Till et al. 2002, p. 57) 

Version 6 of the RSAC program (Wenzel and Schrader 2001) was used extensively for this report to 

provide onsite concentrations due to episodic releases as well as other evaluations. For more 

information on the RSAC program, see Peterson (2004).

Figure 4-2 shows the chronology of facilities and programs at INL. A few comments on the 

development of those facilities will clarify some of the questions that might arise when viewing the INL

map in Figure 4-1 and putting the map into context with the timeline in Figure 4-2.

Figure 4-2. Chronology of facilities and programs at the INL site (DOE 1991). 

EBR-I was the first reactor to operate at the site. It and the Boiling-Water Reactor Experiments 

(BORAX-I through -V), in the southwest corner of the site, were operated under the AEC Chicago 

Operations Office by the University of Chicago as ANL-W. These essentially low-power reactors 

produced very little radioactive airborne effluent (Osloond 1965; Osloond and Newcomb 1969). As 

the EBR-I and BORAX programs were completed, ANL-W relocated to the eastern section of the site

near EBR-II, the TREAT Facility, ZPPR, etc. The EBR-I location is now a National Historic Landmark. 

The Stationary Low-Power Reactor (SL-1), the Mobile Low-Power Reactor No. 1, and the Gas-Cooled 
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 13 of 67

Reactor Experiment (GCRE), which were operated for the U.S. Army, were at the Army (later 

Auxiliary) Reactor Area (ARA). The NRTS Burial Ground became the Radioactive Waste 

Management Complex (RWMC). The short-lived Organic-Moderated Reactor Experiment and 

Experimental Organic-Cooled Reactor (EOCR) were at the EOCR location.

Essentially all of the major facility areas – the TRA, INTEC, TAN, Central Facilities Area (CFA), Naval 

Reactors Facility (NRF), Special Power Excursion Reactor Test (SPERT), and RWMC – have 

operated since the early days of the site. The major changes in these areas are the extent of 

operations at the facilities.

All inhaled quantities and concentrations referred to in this document apply to individuals stationed at 

the site. DOE and INL contractor employees in Idaho Falls, Idaho, in IDO contractor facilities (Willow 

Creek Building, the INL Research Center, the Computer Science Center, etc.) are not affected by site 

activities and, thus, are not subject to inhaled quantities and concentrations [3].

4.2 INTERNAL INTAKES FROM ONSITE AIRBORNE RADIONUCLIDE CONCENTRATION

This section addresses onsite concentrations of radionuclides from normal operational releases and 

from shorter term releases such as those from criticality incidents at INTEC and HTRE No. 3 during 

Initial Engine Test (IET) No. 13, FEBTs, etc. As stated above, operational releases from INTEC and 

TRA have been the predominant and radiologically significant releases at INL during the history of the 

site. For more discussion of these releases and their relationship to other, less significant releases, 

see Peterson (2004), DOE (1991), and Till et al. (2002). 

For worker dose reconstruction, the analyst should use default values for the calculation under 

consideration. When solubility is of concern, use an insoluble oxide form for solids for analysis, with 

type S and M materials being the predominant form [4]. Without more definitive information on the 

type of material, use the material that maximizes the dose for a particular situation. When iodines are 

of concern, consider them to be type F [5].

4.2.1

To determine onsite concentrations of radionuclides from operational releases at INL facility locations, 

the same methodology used to determine offsite concentrations for annual EMRs is used. The 

release for each year of operation is exactly the same as that documented in DOE (1991) with one 

exception: an analysis was performed to reduce the number of radionuclides and yet retain those that 

contributed about 95% of the inhalation dose. This analysis, included in Peterson (2004), reduced the 

number of radionuclides from 56 to 9 for the operational releases. 

Operational Releases

Meteorological dispersion factors applicable to each INL facility were picked from the annual average 

mesoscale dispersion isopleths of ground-level air concentrations published in the EMRs for INL, as 

described in DOE (1991). As described in Appendix B of that document, dispersion isopleths are 

available for the years beginning in 1973, with the exception of 1978 when the telemetry system was 

upgraded. For the years before 1973, the 9-year average of mesoscale dispersion isopleths of 

ground-level air concentrations (DOE 1991) shown in Figure 4-3 was used.2

 2 As used at INL, this quantity is the sum of 8,766 calculations of the hourly average χ/Q.

 For 1993 to 2008, 

annual average mesoscale isopleths from the annual environmental reports (Mitchell 1994; Mitchell, 

Peterson, and Hoff 1995; Mitchell et al. 1996; Mitchell et al. 1997; Evans et al. 1998; Saffel et al. 

2000; Stoller 2002a,b,c, 2004, 2005, 2007, 2008) were used to calculate the facility annual 

concentration. 
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 14 of 67

Of the many facilities on INL, eight facility areas have been chosen for analysis: TAN, INTEC, TRA, 

RWMC, CFA, SPERT, ARA, and ANL-W. TAN includes the IET, Core Test Facility, Technical 

Support Facility (TSF), Loss-of-Fluid Test (LOFT) Facility, Specific Manufacturing Capability, Water 

Figure 4-3. Nine-year (1974 to 1983 excluding 1978) average mesoscale dispersion 

isopleths of air concentration at ground level (hr2

/m3 × 10-9), normalized to unit release 

rate (DOE 1991). 

Reactor Research Test Facility, Process Experimental Pilot Plant, and Low Power Test Facility

(LPTF), depending on the year of operation (ORAUT 2007). The SPERT area includes the Waste 

Experimental Reduction Facility (WERF) and the Power Burst Facility (PBF) depending on the year of 

operation. Facilities such as GRID III and the Experimental Field Station (EFS) are inhabited 

infrequently and have not been included. These facilities were staffed with personnel who were 

normally employed at CFA. An isopleth chosen for a given year for a particular facility, such as 

SPERT, is assumed to apply to PBF, SPERT-I, SPERT-II, etc. If a facility was between two isopleths, 

the higher valued isopleth was chosen. Yearly isopleth values for each of the eight facilities have 

been extracted from the annual EMRs and converted from the normalized annual concentration2

(hr2

/m3) to concentrations (becquerels per cubic meter) and multiplied by 2.4 × 103 m3/yr (the amount 

of air breathed occupationally per year) to produce activity inhaled per year (becquerels) for an 

occupational individual. These are listed in Tables 4-1 through 4-8 for each of the facility areas.

The annual inhaled quantities (becquerels per year) listed in Tables 4-1 through 4-8 are based on 

known and published annual airborne emissions. The 2006 through 2008 values are the 95th
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 15 of 67

percentile of the average of the most recent 10 years of data. In order to address thyroid cancers, I129 environmental intakes have been added. Iodine-129 values were calculated by multiplying the 

95th percentile ratio of 129I to 90Sr stack release data (1,800) by the 90Sr intake. The following 

discussion provides information found in site documentation about facility environmental sampling and 

monitoring and provides data that can be compared with these calculations. 
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 16 of 67

Table4-1. Intake (Bq/yr) by year for ANL-W, 1952 to 2008 [6].

Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, -240 Ru-106 Sr-89 Sr-90Y-91

I-129

1952 5.5E+0 3.3E-2 0.0E+0 0.0E+0 0.0E+0 4.1E-1 0.0E+0 1.3E-1 5.0E-1 2.3E+02

1953 5.5E+0 7.3E-2 1.3E+0 8.9E-4 1.3E-4 4.1E-1 4.0E-1 6.8E-1 8.6E-1 1.2E+03

1954 1.4E+1 5.3E-2 3.3E+0 2.2E-3 3.3E-4 1.0E+0 9.9E-1 1.2E+0 2.1E+0 2.2E+03

1955 1.9E+1 7.9E-2 4.7E+0 2.2E-3 3.3E-4 1.4E+0 1.4E+0 1.7E+0 3.0E+0 3.1E+03

1956 2.2E+1 9.4E-1 5.4E+0 3.6E-3 5.3E-4 1.6E+0 1.7E+0 1.8E+0 3.5E+0 3.2E+03

1957 4.4E+0 1.1E+2 1.1E+1 6.3E-3 9.3E-4 4.7E-1 1.6E+0 2.9E+0 1.7E+0 5.2E+03

1958 6.0E+0 8.3E+1 1.6E+1 9.1E-3 1.3E-3 6.6E-1 1.2E+0 4.1E+0 1.3E+0 7.4E+03

1959 4.8E+01.8E+1 1.3E+1 7.3E-3 1.1E-3 5.4E-1 6.1E-1 3.4E+0 6.7E-1 6.1E+03

1960 6.6E-2 2.6E+0 1.4E-1 2.5E-3 3.7E-4 6.6E-3 7.6E-2 2.1E-1 8.2E-2 3.8E+02

1961 4.6E-2 3.5E+0 3.3E-3 4.6E-4 6.8E-5 2.8E-3 2.0E-1 2.5E-1 2.1E-1 4.5E+02

1962 2.1E-1 3.3E+0 4.6E-1 3.0E-4 4.5E-5 2.1E-2 2.0E-1 3.8E-1 2.2E-1 6.8E+02

1963 3.3E+0 2.1E+0 9.4E+0 2.8E-3 4.2E-4 3.7E-1 1.3E-1 2.8E+0 1.4E-1 5.0E+03

1964 1.8E+0 1.1E-1 0.0E+0 1.1E-4 1.6E-5 2.8E+1 3.0E-2 7.1E-1 2.0E+0 1.3E+03

1965 4.6E+0 7.4E-1 0.0E+0 4.7E-3 7.0E-4 2.0E+0 0.0E+0 2.7E+0 1.8E+0 4.9E+03

1966 2.8E+0 4.3E-1 0.0E+0 1.1E-3 1.6E-4 1.2E+1 0.0E+0 7.8E-1 1.2E+0 1.4E+03

1967 7.0E-2 1.8E-1 0.0E+0 1.2E-4 1.8E-5 1.7E+0 0.0E+0 2.1E-1 6.6E-1 3.8E+02

1968 5.0E+0 3.4E-1 0.0E+0 2.3E-3 3.4E-4 7.4E-1 0.0E+0 1.2E+0 6.4E-1 2.2E+03

1969 2.8E-1 4.8E-1 0.0E+0 5.0E-4 7.4E-5 3.7E-1 0.0E+0 3.6E-1 5.3E-1 6.5E+02

1970 6.6E-1 2.5E-5 0.0E+0 7.0E-4 1.1E-4 3.0E-1 0.0E+0 2.7E-1 5.5E-1 4.9E+02

1971 2.5E+0 7.0E-1 0.0E+0 2.1E-3 3.1E-4 3.2E+0 0.0E+0 1.1E+0 4.7E-1 2.0E+03

1972 2.7E-1 2.9E-1 0.0E+0 6.5E-4 9.7E-5 4.9E-1 0.0E+0 2.8E-1 1.7E-1 5.0E+02

1973 2.4E-2 1.4E-5 0.0E+0 2.4E-4 3.5E-5 1.1E-1 0.0E+0 6.2E-2 2.2E-2 1.1E+02

1974 7.9E-3 1.3E-3 0.0E+0 1.1E-4 9.7E-6 4.4E-2 0.0E+0 3.7E-2 1.2E-1 6.7E+01

1975 8.9E-3 4.5E-3 0.0E+0 1.3E-4 2.5E-5 6.4E-2 0.0E+0 1.9E-2 2.5E-1 3.4E+01

1976 5.2E-5 3.2E-5 0.0E+0 8.1E-6 3.6E-6 8.1E-4 0.0E+0 3.9E-4 3.0E-2 7.0E

-01

1977 2.0E-4 1.4E-4 0.0E+0 8.0E-5 3.4E-5 1.1E-2 0.0E+0 5.0E-3 4.3E-1 9.0E+00

1978 3.8E-4 2.0E-3 0.0E+0 7.4E-5 7.9E-6 5.7E-3 0.0E+0 1.9E-3 3.5E-1 3.4E+00

1979 1.8E-4 9.7E-5 0.0E+0 4.8E-5 5.2E-6 1.3E-3 0.0E+0 8.9E-3 5.1E-2 1.6E+01

1980 2.9E-4 1.5E-3 0.0E+0 3.1E-5 4.0E-6 6.3E-4 0.0E+0 4.3E-4 3.1E-1 7.7E

-01

1981 2.9E-4 3.8E-3 0.0E+0 6.1E-6 1.1E-6 6.1E-3 0.0E+0 3.3E-4 2.0E-1 5.9E

-01

1982 1.5E-4 4.7E-5 0.0E+0 1.5E-51.6E-6 4.4E-4 0.0E+0 2.8E-4 7.2E-2 5.0E

-01

1983 2.9E-4 1.5E-3 0.0E+0 1.2E-4 1.6E-5 2.3E-3 0.0E+0 1.1E-4 3.6E-2 2.0E

-01

1984 2.9E-4 9.7E-5 0.0E+0 1.9E-5 7.4E-6 3.2E-4 0.0E+0 1.3E-4 1.4E-2 2.3E

-01

1985 1.2E-3 9.0E-3 0.0E+0 2.3E-5 4.5E-6 1.0E-2 0.0E+0 6.6E-4 9.9E-1 1.2E+00

1986 2.9E-4 8.9E-5 0.0E+0 1.3E-6 9.7E-8 2.3E-3 0.0E+0 1.6E-5 4.3E-2 2.9E

-02

1987 2.9E-4 4.3E-5 0.0E+0 1.4E-6 2.1E-7 3.0E-5 0.0E+0 2.3E-5 7.6E-1 4.1E

-02

1988 2.9E-4 1.4E-5 0.0E+0 1.1E-6 1.7E-7 1.5E-2 0.0E+0 2.7E-5 5.2E-1 4.9E

-02

1989 2.9E-4 9.7E-6 0.0E+0 5.7E-9 8.1E-10 1.6E-4 0.0E+0 7.4E-6 6.7E-2 1.3E

-02
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 17 of 67

Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, -240 Ru-106 Sr-89 Sr-90Y-91

I-129

1990 1.2E-4 3.8E-5 0.0E+0 9.4E-10 9.4E-10 4.2E-5 0.0E+0 2.0E-7 3.1E-2 3.6E

-04

1991 1.2E-4 1.6E-5 0.0E+0 1.0E-10 1.0E-10 5.2E-5 0.0E+0 9.7E-5 1.8E-2 1.7E

-01

1992 1.2E-4 1.8E-5 0.0E+01.7E-7 1.7E-7 1.4E-5 0.0E+0 8.3E-6 3.1E-2 1.5E

-02

1993 0.0E+0 3.8E-6 0.0E+0 0.0E+0 8.7E-11 3.5E-5 0.0E+0 3.8E-5 0.0E+0 6.8E

-02

1994 0.0E+0 3.1E-5 0.0E+0 0.0E+0 4.6E-8 0.0E+0 0.0E+0 8.0E-5 0.0E+0 1.4E

-01

1995 0.0E+0 2.1E-5 0.0E+0 3.3E-8 5.5E-9 0.0E+0 0.0E+0 2.3E-6 0.0E+0 4.1E

-03

1996 0.0E+0 2.8E-5 0.0E+0 2.2E-7 4.5E-9 0.0E+0 0.0E+0 1.1E-6 0.0E+0 2.0E

-03

1997 0.0E+0 0.0E+0 0.0E+0 1.8E-7 5.5E-8 0.0E+0 0.0E+0 2.4E-5 0.0E+0 4.3E

-02

1998 0.0E+0 2.3E-5 0.0E+0 1.7E-7 1.8E-8 0.0E+0 0.0E+0 1.1E-5 0.0E+0 2.0E

-02

1999 0.0E+0 3.1E-5 0.0E+0 7.5E-8 7.1E-9 0.0E+0 0.0E+0 4.4E-6 0.0E+0 7.9E

-03

2000 0.0E+0 1.9E-3 0.0E+0 3.6E-5 3.6E-7 0.0E+0 0.0E+0 3.5E-3 0.0E+0 6.3E+00

2001 0.0E+0 1.0E-3 0.0E+0 2.4E-7 2.9E-5 0.0E+0 0.0E+0 1.2E-4 0.0E+0 2.2E

-01

2002 0.0E+0 1.6E-4 0.0E+0 6.6E-6 1.8E-5 0.0E+0 0.0E+0 3.8E-3 0.0E+0 6.8E+00

2003 9.4E-8 1.2E-2 2.4E-2 1.0E-5 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+00

2004 1.8E-5 1.2E-4 4.9E-4 4.0E-5 5.6E-5 0.0E+0 5.9E-6 1.1E-3 0.0E+0 2.0E+00

2005 8.8E-7 2.3E-2 8.4E-5 4.5E-5 2.6E-4 4.8E-8 3.8E-4 1.3E-2 0.0E+0 2.3E+01

2006–2008 3.19E-6 4.36E-3 4.27E-3 1.08E-5 4.57E-5 8.55E-9 6.76E-5 2.32E-3 0.00E+0 4.2E+00

Table4-2. Intake (Bq/yr) by year for ARA, 1952 to 2008 [6].

Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90Y-91

I-129

1952 7.9E+0 4.7E-2 0.0E+0 0.0E+0 0.0E+0 5.8E-1 0.0E+0 1.8E-1 7.2E-1 3.24E+02

1953 7.9E+0 1.0E-1 1.9E+0 1.3E-3 1.9E-4 5.8E-1 5.7E-1 9.7E-1 1.2E+0 1.75E+03

1954 2.0E+1 7.6E-2 4.8E+0 3.2E-3 4.7E-4 1.5E+0 1.4E+0 1.7E+0 3.1E+0 3.06E+03

1955 2.8E+1 1.1E-1 6.7E+0 3.2E-3 4.7E-4 2.0E+0 2.0E+0 2.4E+0 4.3E+0 4.32E+03

1956 3.2E+1 1.3E+0 7.7E+0 5.1E-3 7.5E-4 2.3E+0 2.4E+0 2.5E+0 5.0E+0 4.50E+03

1957 6.2E+0 1.6E+2 1.6E+1 9.0E-3 1.3E-3 6.8E-1 2.2E+0 4.1E+0 2.4E+0 7.38E+03

1958 8.5E+0 1.2E+2 2.3E+1 1.3E-2 1.9E-3 9.4E-1 1.7E+0 5.9E+0 1.9E+0 1.06E+04

1959 6.9E+0 2.6E+1 1.9E+1 1.0E-2 1.5E-3 7.7E-1 8.7E-1 4.9E+0 9.6E-1 8.82E+03

1960 9.5E-2 3.8E+0 2.0E-1 3.6E-3 5.3E-4 9.4E-3 1.1E-1 3.0E-1 1.2E-1 5.40E+02

1961 6.6E-2 4.9E+0 4.7E-3 6.6E-4 9.7E-5 3.9E-3 2.8E-1 3.6E-1 3.1E-1 6.48E+02

1962 3.0E-1 4.7E+0 6.5E-1 4.3E-4 6.4E-5 3.0E-2 2.9E-1 5.4E-1 3.1E-1 9.72E+02

1963 4.8E+0 3.1E+0 1.3E+1 4.1E-3 5.9E-4 5.4E-1 1.8E-1 4.0E+0 2.0E-1 7.20E+03

1964 2.6E+0 1.6E-1 0.0E+0 1.6E-4 2.3E-5 3.9E+1 4.3E-2 1.0E+0 2.8E+0 1.80E+03

1965 6.5E+0 1.1E+0 0.0E+0 6.8E-3 1.0E-3 2.8E+0 0.0E+0 3.9E+0 2.5E+0 7.02E+03

1966 3.9E+0 6.2E-1 0.0E+0 1.5E-3 2.3E-4 1.8E+1 0.0E+0 1.1E+0 1.7E+0 1.98E+03

1967 9.9E-2 2.6E-1 0.0E+0 1.7E-4 2.5E-5 2.5E+0 0.0E+0 3.0E-1 9.5E-1 5.40E+02

1968 7.1E+0 4.8E-1 0.0E+0 3.3E-3 4.8E-4 1.1E+0 0.0E+0 1.7E+0 9.1E-1 3.06E+03

1969 4.0E-1 6.9E-1 0.0E+0 7.2E-4 1.1E-4 5.3E-1 0.0E+0 5.1E-1 7.6E-1 9.18E+02

1970 9.5E-1 3.6E-5 0.0E+0 1.0E-3 1.5E-4 4.3E-1 0.0E+0 3.8E-1 7.8E-1 6.84E+02
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 18 of 67

Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90Y-91

I-129

1971 3.6E+0 9.9E-1 0.0E+0 3.0E-3 4.4E-4 4.6E+0 0.0E+0 1.6E+0 6.6E-1 2.88E+03

1972 3.9E-1 4.2E-1 0.0E+0 9.2E-4 1.4E-4 7.0E-1 0.0E+0 4.0E-1 2.4E-1 7.20E+02

1973 8.1E-3 4.6E-6 0.0E+0 8.0E-5 1.2E-5 3.8E-2 0.0E+0 2.1E-2 7.3E-3 3.78E+01

1974 2.4E-2 3.8E-3 0.0E+0 3.3E-4 2.9E-5 1.3E-1 0.0E+0 1.1E-1 3.7E-1 1.98E+02

1975 1.3E-2 6.5E-3 0.0E+0 1.8E-4 3.6E-5 9.1E-2 0.0E+0 2.8E-2 3.5E-1 5.04E+01

1976 7.4E-5 4.6E-5 0.0E+0 1.2E-5 5.1E-6 1.2E-3 0.0E+0 5.5E-4 4.3E-2 9.90E

-01

1977 2.9E-4 2.0E-4 0.0E+0 1.1E-44.8E-5 1.5E-2 0.0E+0 7.2E-3 6.1E-1 1.30E+01

1978 5.4E-4 2.9E-3 0.0E+0 1.1E-4 1.1E-5 8.1E-3 0.0E+0 2.8E-3 5.0E-1 5.04E+00

1979 7.6E-4 4.2E-4 0.0E+0 2.0E-4 2.2E-5 5.5E-3 0.0E+0 3.8E-2 2.2E-1 6.84E+01

1980 1.2E-3 6.2E-3 0.0E+0 1.3E-4 1.7E-5 2.7E-3 0.0E+0 1.8E-3 1.3E+0 3.24E+00

1981 1.2E-3 1.6E-2 0.0E+0 2.6E-5 4.8E-6 2.6E-2 0.0E+0 1.4E-3 8.6E-1 2.52E+00

1982 1.5E-4 4.7E-5 0.0E+0 1.5E-5 1.6E-6 4.4E-4 0.0E+0 2.8E-4 7.2E-2 5.04E

-01

1983 4.1E-4 2.1E-3 0.0E+0 1.7E-4 2.3E-5 3.3E-3 0.0E+0 1.5E-4 5.2E-2 2.70E

-01

1984 4.1E-4 1.4E-4 0.0E+0 2.7E-5 1.1E-5 4.5E-4 0.0E+0 1.8E-4 2.1E-2 3.24E

-01

1985 1.2E-3 9.0E-3 0.0E+0 2.3E-5 4.5E-6 1.0E-2 0.0E+0 6.6E-4 9.9E-1 1.19E+00

1986 1.2E-3 3.8E-4 0.0E+0 5.5E-6 4.2E-7 1.0E-2 0.0E+0 6.9E-5 1.9E-1 1.24E

-01

1987 2.9E-4 4.3E-5 0.0E+0 1.4E-6 2.1E-7 3.0E-5 0.0E+0 2.3E-5 7.6E-1 4.14E

-02

1988 2.9E-4 1.4E-5 0.0E+0 1.1E-6 1.7E-7 1.5E-2 0.0E+0 2.7E-5 5.2E-1 4.86E

-02

1989 1.2E-3 4.2E-5 0.0E+0 2.5E-8 3.5E-9 6.9E-4 0.0E+0 3.2E-5 2.9E-1 5.76E

-02

1990 4.1E-4 1.3E-4 0.0E+0 3.1E-9 3.1E-9 1.4E-4 0.0E+0 6.6E-7 1.0E-1 1.19E

-03

1991 4.1E-4 5.2E-5 0.0E+0 3.3E-10 3.3E-10 1.7E-4 0.0E+0 3.2E-4 6.0E-2 5.76E

-01

1992 4.1E-4 6.0E-5 0.0E+0 5.7E-7 5.7E-7 4.6E-5 0.0E+0 2.8E-5 1.0E-1 5.04E

-02

1993 0.0E+0 3.8E-5 0.0E+0 0.0E+0 8.7E-10 3.5E-4 0.0E+0 3.8E-4 0.0E+0 6.84E

-01

1994 0.0E+0 1.3E-4 0.0E+0 0.0E+0 2.0E-7 0.0E+0 0.0E+0 3.4E-4 0.0E+0 6.12E

-01

1995 0.0E+0 4.8E-5 0.0E+0 7.7E-8 1.3E-8 0.0E+0 0.0E+0 5.4E-6 0.0E+0 9.72E

-03

1996 0.0E+0 6.5E-5 0.0E+0 5.1E-7 1.0E-8 0.0E+0 0.0E+0 2.5E-6 0.0E+0 4.50E

-03

1997 0.0E+0 0.0E+0 0.0E+0 4.1E-7 1.3E-7 0.0E+0 0.0E+0 5.7E-5 0.0E+0 1.03E

-01

1998 0.0E+0 5.4E-5 0.0E+0 4.0E-7 4.3E-8 0.0E+0 0.0E+0 2.5E-5 0.0E+0 4.50E

-02

1999 0.0E+0 1.0E-4 0.0E+0 2.5E-7 2.4E-8 0.0E+0 0.0E+0 1.5E-5 0.0E+0 2.70E

-02

2000 0.0E+0 6.4E-3 0.0E+0 1.2E-4 1.2E-6 0.0E+0 0.0E+0 1.2E-2 0.0E+0 2.16E+01

2001 0.0E+0 3.3E-3 0.0E+0 8.0E-7 9.5E-5 0.0E+0 0.0E+0 3.9E-4 0.0E+0 7.02E

-01

2002 0.0E+0 3.7E-4 0.0E+0 1.6E-5 4.2E-5 0.0E+0 0.0E+0 8.9E-3 0.0E+0 1.60E+01

2003 5.7E

-8 7.2E-3 1.4E-2 6.3E-6 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.00E+00

2004 6.9E

-6 4.4E-5 1.8E-4 1.5E-5 2.1E-5 0.0E+0 2.2E-6 4.0E-4 0.0E+0 7.20E

-01

2005 8.8E

-7 2.3E-2 8.4E-5 4.5E-5 2.6E-4 4.8E-8 3.8E-4 1.3E-2 0.0E+0 2.34E+01

2006–2008 1.22E-6 4.11E-3 2.49E-3 2.14E-5 4.68E-5 8.55E-9 6.77E-5 3.14E-3 0.00E+0 5.65E+00
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 19 of 67

Table4-3. Intake (Bq/yr) by year for CFA, 1952 to 2008 [6].

Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90Y-91

I-129

1952 2.4E+1 1.4E-1 0.0E+0 0.0E+00.0E+0 1.7E+0 0.0E+0 5.4E-1 2.2E+0 9.7E+02

1953 2.4E+1 3.1E-1 5.7E+0 3.8E-3 5.6E-4 1.7E+0 1.7E+0 2.9E+0 3.7E+0 5.2E+03

1954 6.0E+1 2.3E-1 1.4E+1 9.5E-3 1.4E-3 4.4E+0 4.3E+0 5.1E+0 9.2E+0 9.2E+03

1955 8.3E+1 3.4E-1 2.0E+1 9.5E-3 1.4E-3 6.1E+0 6.0E+0 7.3E+0 1.3E+1 1.3E+04

1956 9.6E+1 4.0E+0 2.3E+1 1.5E-2 2.3E-3 7.0E+0 7.2E+0 7.6E+0 1.5E+1 1.4E+04

1957 1.9E+1 4.8E+2 4.8E+1 2.7E-2 4.0E-3 2.0E+0 6.6E+0 1.2E+1 7.3E+0 2.2E+04

1958 2.6E+1 3.6E+2 6.9E+1 3.9E-2 5.7E-3 2.8E+0 5.1E+0 1.8E+1 5.6E+0 3.2E+04

1959 2.1E+1 7.8E+1 5.7E+1 3.1E-2 4.6E-3 2.3E+0 2.6E+0 1.5E+1 2.9E+0 2.7E+04

1960 2.8E-1 1.1E+1 6.0E-1 1.1E-2 1.6E-3 2.8E-2 3.2E-1 9.0E-1 3.5E-1 1.6E+03

1961 2.0E-1 1.5E+1 1.4E-2 2.0E-3 2.9E-4 1.2E-2 8.5E-1 1.1E+0 9.2E-1 2.0E+03

1962 8.9E-1 1.4E+1 2.0E+0 1.3E-3 1.9E-4 8.9E-2 8.6E-1 1.6E+0 9.3E-1 2.9E+03

1963 1.4E+1 9.2E+0 4.0E+1 1.2E-2 1.8E-3 1.6E+0 5.4E-1 1.2E+1 6.0E-1 2.2E+04

1964 7.8E+0 4.8E-1 0.0E+0 4.7E-4 6.9E-5 1.2E+2 1.3E-1 3.0E+0 8.5E+0 5.4E+03

1965 2.0E+1 3.2E+0 0.0E+0 2.0E-2 3.0E-3 8.5E+0 0.0E+0 1.2E+1 7.5E+0 2.2E+04

1966 1.2E+1 1.9E+0 0.0E+0 4.6E-3 6.8E-4 5.3E+1 0.0E+0 3.3E+0 5.0E+0 5.9E+03

1967 3.0E-1 7.9E-1 0.0E+0 5.2E-4 7.6E-5 7.4E+0 0.0E+0 9.0E-1 2.8E+0 1.6E+03

1968 2.1E+1 1.4E+0 0.0E+0 1.0E-2 1.5E-3 3.2E+0 0.0E+0 5.2E+0 2.7E+0 9.4E+03

1969 1.2E+0 2.1E+0 0.0E+0 2.1E-3 3.2E-4 1.6E+0 0.0E+0 1.5E+0 2.3E+0 2.7E+03

1970 2.8E+0 1.1E-4 0.0E+0 3.0E-3 4.5E-4 1.3E+0 0.0E+0 1.1E+0 2.3E+0 2.0E+03

1971 1.1E+1 3.0E+0 0.0E+0 9.0E-3 1.3E-3 1.4E+1 0.0E+0 4.8E+0 2.0E+0 8.6E+03

1972 2.7E-1 2.9E-1 0.0E+0 6.5E-4 9.7E-5 4.9E-1 0.0E+0 2.8E-1 1.7E-1 5.0E+02

1973 2.4E-2 1.4E-5 0.0E+0 2.4E-4 3.5E-5 1.1E-1 0.0E+0 6.2E-2 2.2E-2 1.1E+02

1974 7.9E-3 1.3E-3 0.0E+0 1.1E-4 9.7E-6 4.4E-2 0.0E+0 3.7E-2 1.2E-1 6.7E+01

1975 8.9E-3 4.5E-3 0.0E+0 1.3E-4 2.5E-5 6.4E-2 0.0E+01.9E-2 2.5E-1 3.4E+01

1976 5.2E-5 3.2E-5 0.0E+0 8.1E-6 3.6E-6 8.1E-4 0.0E+0 3.9E-4 3.0E-2 7.0E

-01

1977 2.0E-4 1.4E-4 0.0E+0 8.0E-5 3.4E-5 1.1E-2 0.0E+0 5.0E-3 4.3E-1 9.0E+00

1978 3.8E-4 2.0E-3 0.0E+0 7.4E-5 7.9E-6 5.7E-3 0.0E+0 1.9E-3 3.5E-1 3.4E+00

1979 1.8E-4 9.7E-5 0.0E+0 4.8E-5 5.2E-6 1.3E-3 0.0E+0 8.9E-3 5.1E-2 1.6E+01

1980 2.9E-4 1.5E-3 0.0E+0 3.1E-5 4.0E-6 6.3E-4 0.0E+0 4.3E-4 3.1E-1 7.7E

-01

1981 2.9E-4 3.8E-3 0.0E+0 6.1E-6 1.1E-6 6.1E-3 0.0E+0 3.3E-4 2.0E-1 5.9E

-01

1982 1.5E-4 4.7E-5 0.0E+0 1.5E-5 1.6E-6 4.4E-4 0.0E+0 2.8E-4 7.2E-2 5.0E

-01

1983 2.9E-4 1.5E-3 0.0E+0 1.2E-4 1.6E-5 2.3E-3 0.0E+0 1.1E-4 3.6E-2 2.0E

-01

1984 2.9E-4 9.7E-5 0.0E+0 1.9E-5 7.4E-6 3.2E-4 0.0E+0 1.3E-4 1.4E-2 2.3E

-01

1985 1.2E-3 9.0E-3 0.0E+0 2.3E-5 4.5E-6 1.0E-2 0.0E+0 6.6E-4 9.9E-1 1.2E+00

1986 2.9E-4 8.9E-5 0.0E+0 1.3E-6 9.7E-8 2.3E-3 0.0E+0 1.6E-5 4.3E-2 2.9E

-02

1987 2.9E-4 4.3E-5 0.0E+0 1.4E-6 2.1E-7 3.0E-5 0.0E+0 2.3E-5 7.6E-1 4.1E

-02

1988 2.9E-4 1.4E-5 0.0E+0 1.1E-6 1.7E-7 1.5E-2 0.0E+0 2.7E-5 5.2E-1 4.9E

-02

1989 2.9E-4 9.7E-6 0.0E+0 5.7E-9 8.1E-10 1.6E-4 0.0E+0 7.4E-6 6.7E-2 1.3E

-02
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 20 of 67

Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90Y-91

I-129

1990 1.2E-4 3.8E-5 0.0E+0 9.4E-10 9.4E-10 4.2E-5 0.0E+0 2.0E-7 3.1E-2 3.6E

-04

1991 1.2E-4 1.6E-5 0.0E+0 1.0E-10 1.0E-10 5.2E-5 0.0E+0 9.7E-5 1.8E-2 1.7E

-01

1992 1.2E-4 1.8E-5 0.0E+0 1.7E-7 1.7E-7 1.4E-5 0.0E+0 8.3E-6 3.1E-2 1.5E

-02

1993 0.0E+0 3.8E-6 0.0E+0 0.0E+0 8.7E-11 3.5E-5 0.0E+0 3.8E-5 0.0E+0 6.8E

-02

1994 0.0E+0 3.1E-5 0.0E+0 0.0E+0 4.6E-8 0.0E+0 0.0E+0 8.0E-5 0.0E+0 1.4E

-01

1995 0.0E+0 2.1E-5 0.0E+0 3.3E-8 5.5E-9 0.0E+0 0.0E+0 2.3E-6 0.0E+0 4.1E

-03

1996 0.0E+0 2.8E-5 0.0E+0 2.2E-7 4.5E-9 0.0E+0 0.0E+0 1.1E-6 0.0E+0 2.0E

-03

1997 0.0E+0 0.0E+0 0.0E+0 1.8E-7 5.5E-8 0.0E+0 0.0E+0 2.4E-5 0.0E+0 4.3E

-02

1998 0.0E+0 2.3E-5 0.0E+0 1.7E-7 1.8E-8 0.0E+0 0.0E+0 1.1E-5 0.0E+0 2.0E

-02

1999 0.0E+0 3.1E-5 0.0E+0 7.5E-8 7.1E-9 0.0E+0 0.0E+0 4.4E-6 0.0E+0 7.9E

-03

2000 0.0E+0 1.9E-3 0.0E+0 3.6E-5 3.6E-7 0.0E+0 0.0E+0 3.5E-3 0.0E+0 6.3E+00

2001 0.0E+0 1.0E-3 0.0E+0 2.4E-7 2.9E-5 0.0E+0 0.0E+0 1.2E-4 0.0E+0 2.2E

-01

2002 0.0E+0 5.3E-4 0.0E+0 2.2E-5 6.0E-5 0.0E+0 0.0E+0 1.3E-2 0.0E+0 2.3E+01

2003 3.8E

-7 4.8E-2 9.5E-2 4.2E-5 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+00

2004 2.8E

-5 1.8E-4 7.3E-4 6.0E-5 8.4E-5 0.0E+0 8.9E-6 1.6E-3 0.0E+0 2.9E+00

2005 1.9E

-6 4.9E-2 1.8E-4 9.6E-5 5.6E-4 1.0E-7 8.1E-4 2.9E-2 0.0E+0 5.2E+01

2006–2008 4.96E

-6 1.15E-2 1.69E-2 1.90E-5 9.82E-5 1.78E-8 1.44E-4 5.36E-3 0.00E+0 9.6E+00

Table4-4. Intake (Bq/yr) by year for INTEC, 1952 to 2008 [6].

Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90Y-91

I-129

1952 2.4E+1 1.4E-1 0.0E+0 0.0E+0 0.0E+0 1.7E+0 0.0E+0 5.4E-1 2.2E+0 9.7E+02

1953 2.4E+1 3.1E-1 5.7E+0 3.8E-3 5.6E-4 1.7E+0 4.0E-1 2.9E+0 3.7E+0 5.2E+03

1954 6.0E+1 2.3E-1 1.4E+1 9.5E-3 1.4E-3 4.4E+0 9.9E-1 5.1E+0 9.2E+0 9.2E+03

1955 8.3E+1 3.4E-1 2.0E+1 9.5E-3 1.4E-3 6.1E+0 1.4E+0 7.3E+0 1.3E+1 1.3E+04

1956 9.6E+1 4.0E+0 2.3E+1 1.5E-2 2.3E-3 7.0E+0 1.7E+0 7.6E+0 1.5E+1 1.4E+04

1957 1.9E+1 4.8E+2 4.8E+1 2.7E-2 4.0E-3 2.0E+0 1.6E+0 1.2E+1 7.3E+0 2.2E+04

1958 2.6E+1 3.6E+2 6.9E+1 3.9E-2 5.7E-3 2.8E+0 1.2E+0 1.8E+1 5.6E+0 3.2E+04

1959 2.1E+1 7.8E+1 5.7E+1 3.1E-2 4.6E-3 2.3E+0 6.1E-1 1.5E+1 2.9E+0 2.7E+04

1960 2.8E-1 1.1E+1 6.0E-1 1.1E-2 1.6E-3 2.8E-2 7.6E-2 9.0E-1 3.5E-1 1.6E+03

1961 2.0E-1 1.5E+1 1.4E-2 2.0E-3 2.9E-4 1.2E-2 2.0E-1 1.1E+0 9.2E-1 2.0E+03

1962 8.9E-1 1.4E+1 2.0E+0 1.3E-3 1.9E-4 8.9E-2 2.0E-1 1.6E+0 9.3E-1 2.9E+03

1963 1.4E+1 9.2E+0 4.0E+1 1.2E-2 1.8E-3 1.6E+0 1.3E-1 1.2E+1 6.0E-1 2.2E+04

1964 7.8E+0 4.8E-1 0.0E+0 4.7E-4 6.9E-5 1.2E+2 3.0E-2 3.0E+0 8.5E+0 5.4E+03

1965 2.0E+1 3.2E+0 0.0E+0 2.0E-2 3.0E-3 8.5E+0 0.0E+0 1.2E+1 7.5E+0 2.2E+04

1966 1.2E+1 1.9E+0 0.0E+0 4.6E-3 6.8E-4 5.3E+1 0.0E+0 3.3E+0 5.0E+0 5.9E+03

1967 3.0E-1 7.9E-1 0.0E+0 5.2E-4 7.6E-5 7.4E+0 0.0E+0 9.0E-1 2.8E+0 1.6E+03

1968 2.1E+1 1.4E+0 0.0E+0 1.0E-2 1.5E-3 3.2E+0 0.0E+0 5.2E+0 2.7E+0 9.4E+03

1969 1.2E+0 2.1E+0 0.0E+0 2.1E-3 3.2E-4 1.6E+0 0.0E+0 1.5E+0 2.3E+0 2.7E+03

1970 2.8E+0 1.1E-4 0.0E+0 3.0E-3 4.5E-4 1.3E+0 0.0E+0 1.1E+0 2.3E+0 2.0E+03
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 21 of 67

Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90Y-91

I-129

1971 1.1E+1 3.0E+0 0.0E+0 9.0E-3 1.3E-3 1.4E+1 0.0E+0 4.8E+0 2.0E+0 8.6E+03

1972 1.2E+0 1.2E+0 0.0E+0 2.8E-3 4.2E-4 2.1E+0 0.0E+0 1.2E+0 7.2E-1 2.2E+03

1973 5.7E-2 3.2E-5 0.0E+0 5.6E-4 8.1E-5 2.7E-1 0.0E+0 1.5E-1 5.1E-2 2.7E+02

1974 7.9E-2 1.3E-2 0.0E+0 1.1E-3 9.7E-5 4.4E-1 0.0E+0 3.7E-1 1.2E+0 6.7E+02

1975 3.8E-2 1.9E-2 0.0E+0 5.5E-4 1.1E-4 2.7E-1 0.0E+0 8.3E-2 1.1E+0 1.5E+02

1976 5.2E-4 3.2E-4 0.0E+0 8.1E-5 3.6E-5 8.1E-3 0.0E+0 3.9E-3 3.0E-1 7.0E+00

1977 8.7E-4 5.9E-4 0.0E+0 3.4E-4 1.5E-4 4.5E-2 0.0E+0 2.1E-2 1.8E+0 3.8E+01

1978 1.6E-3 8.7E-3 0.0E+0 3.2E-4 3.4E-5 2.4E-2 0.0E+0 8.3E-3 1.5E+0 1.5E+01

1979 7.6E-4 4.2E-4 0.0E+0 2.0E-4 2.2E-5 5.5E-3 0.0E+0 3.8E-2 2.2E-1 6.8E+01

1980 1.2E-3 6.2E-3 0.0E+0 1.3E-4 1.7E-5 2.7E-3 0.0E+0 1.8E-3 1.3E+0 3.2E+0

0

1981 1.2E-3 1.6E-2 0.0E+0 2.6E-5 4.8E-6 2.6E-2 0.0E+0 1.4E-3 8.6E-1 2.5E+00

1982 2.2E-4 6.7E-5 0.0E+0 2.1E-5 2.3E-6 6.3E-4 0.0E+0 4.0E-4 1.0E-1 7.2E

-01

1983 1.2E-3 6.2E-3 0.0E+0 5.2E-4 6.9E-5 1.0E-2 0.0E+0 4.5E-4 1.5E-1 8.1E

-01

1984 1.2E-3 4.2E-4 0.0E+0 8.0E-5 3.2E-5 1.4E-3 0.0E+0 5.5E-4 6.2E-2 9.9E

-01

1985 2.9E-3 2.1E-2 0.0E+0 5.3E-5 1.1E-5 2.3E-2 0.0E+0 1.5E-3 2.3E+0 2.7E+00

1986 1.2E-3 3.8E-4 0.0E+0 5.5E-6 4.2E-7 1.0E-2 0.0E+0 6.9E-5 1.9E-1 1.2E

-01

1987 1.2E-3 1.8E-4 0.0E+0 5.9E-6 9.0E-7 1.3E-4 0.0E+0 9.7E-5 3.3E+0 1.7E

-01

1988 1.2E-3 5.9E-5 0.0E+0 4.8E-6 7.3E-7 6.6E-2 0.0E+0 1.2E-4 2.2E+0 2.2E

-01

1989 1.2E-3 4.2E-5 0.0E+0 2.5E-8 3.5E-9 6.9E-4 0.0E+0 3.2E-5 2.9E-1 5.8E

-02

1990 4.1E-4 1.3E-4 0.0E+0 3.1E-9 3.1E-9 1.4E-4 0.0E+0 6.6E-7 1.0E-1 1.2E

-03

1991 4.1E-4 5.2E-5 0.0E+0 3.3E-10 3.3E-10 1.7E-4 0.0E+0 3.2E-4 6.0E-2 5.8E

-01

1992 1.2E-3 1.8E-4 0.0E+0 1.7E-6 1.7E-6 1.4E-4 0.0E+0 8.3E-5 3.1E-1 1.5E

-01

1993 0.0E+0 3.8E-5 0.0E+0 0.0E+0 8.7E-10 3.5E-4 0.0E+0 3.8E-4 0.0E+0 6.8E

-01

1994 0.0E+0 1.3E-4 0.0E+0 0.0E+0 2.0E-7 0.0E+0 0.0E+0 3.4E-4 0.0E+0 6.1E

-01

1995 0.0E+0 6.9E-5 0.0E+0 1.1E-7 1.8E-8 0.0E+0 0.0E+0 7.7E-6 0.0E+0 1.4E

-02

1996 0.0E+0 9.2E-5 0.0E+0 7.3E-7 1.5E-8 0.0E+0 0.0E+0 3.6E-6 0.0E+0 6.5E

-03

1997 0.0E+0 0.0E+0 0.0E+0 5.9E-7 1.8E-7 0.0E+0 0.0E+0 8.1E-5 0.0E+0 1.5E

-01

1998 0.0E+0 2.3E-4 0.0E+0 1.7E-6 1.8E-7 0.0E+0 0.0E+0 1.1E-4 0.0E+0 2.0E

-01

1999 0.0E+0 3.1E-4 0.0E+0 7.5E-7 7.1E-8 0.0E+0 0.0E+0 4.4E-5 0.0E+0 7.9E

-02

2000 0.0E+0 6.4E-3 0.0E+0 1.2E-4 1.2E-6 0.0E+0 0.0E+0 1.2E-2 0.0E+0 2.2E+01

2001 0.0E+0 3.3E-3 0.0E+0 8.0E-7 9.5E-5 0.0E+0 0.0E+0 3.9E-4 0.0E+0 7.0E

-01

2002 0.0E+0 1.6E-3 0.0E+0 6.7E-5 1.8E-4 0.0E+0 0.0E+0 3.8E-2 0.0E+0 6.8E+01

2003 5.7E-7 7.2E-2 1.4E-1 6.3E-5 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+00

2004 3.4E-5 2.2E-4 9.1E-4 7.6E-5 1.0E-4 0.0E+0 1.1E-5 2.0E-3 0.0E+0 3.6E+00

2005 3.8E-6 9.8E-2 3.6E-4 1.9E-4 1.1E-3 2.1E-7 1.6E-3 5.8E-2 0.0E+0 1.0E+02

2006–2008 6.02E-6 2.03E-2 2.49E-2 3.74E-5 1.93E-4 3.74E-8 2.85E-4 1.16E-2 0.00E+0 2.1E+01
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 22 of 67

Table4-5. Intake (Bq/yr) by year for RWMC, 1952 to 2008 [6].

Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90Y-91

I-129

1952 5.5E+0 3.3E-2 0.0E+0 0.0E+0 0.0E+0 4.1E-1 0.0E+0 1.3E-1 5.0E-1 2.3E+02

1953 5.5E+0 7.3E-2 1.3E+0 8.9E-4 1.3E-4 4.1E-1 4.0E-1 6.8E-1 8.6E-1 1.2E+03

1954 1.4E+1 5.3E-2 3.3E+0 2.2E-3 3.3E-4 1.0E+0 9.9E-1 1.2E+0 2.1E+0 2.2E+03

1955 1.9E+1 7.9E-2 4.7E+0 2.2E-3 3.3E-4 1.4E+0 1.4E+0 1.7E+0 3.0E+0 3.1E+03

1956 2.2E+1 9.4E-1 5.4E+0 3.6E-3 5.3E-4 1.6E+0 1.7E+0 1.8E+0 3.5E+0 3.2E+03

1957 4.4E+0 1.1E+2 1.1E+1 6.3E-3 9.3E-4 4.7E-1 1.6E+0 2.9E+0 1.7E+0 5.2E+03

1958 6.0E+0 8.3E+1 1.6E+1 9.1E-3 1.3E-3 6.6E-1 1.2E+0 4.1E+0 1.3E+0 7.4E+03

1959 4.8E+0 1.8E+1 1.3E+1 7.3E-3 1.1E-3 5.4E-1 6.1E-1 3.4E+0 6.7E-1 6.1E+03

1960 6.6E-2 2.6E+0 1.4E-1 2.5E-3 3.7E-4 6.6E-3 7.6E-2 2.1E-1 8.2E-2 3.8E+02

1961 4.6E-2 3.5E+0 3.3E-3 4.6E-4 6.8E-5 2.8E-3 2.0E-1 2.5E-1 2.1E-1 4.5E+02

1962 2.1E-1 3.3E+0 4.6E-1 3.0E-4 4.5E-5 2.1E-2 2.0E-1 3.8E-1 2.2E-1 6.8E+02

1963 3.3E+0 2.1E+0 9.4E+0 2.8E-3 4.2E-4 3.7E-1 1.3E-1 2.8E+0 1.4E-1 5.0E+03

1964 1.8E+0 1.1E-1 0.0E+0 1.1E-4 1.6E-5 2.8E+1 3.0E-2 7.1E-1 2.0E+0 1.3E+03

1965 4.6E+0 7.4E-1 0.0E+0 4.7E-3 7.0E-4 2.0E+0 0.0E+0 2.7E+0 1.8E+0 4.9E+03

1966 2.8E+0 4.3E-1 0.0E+0 1.1E-3 1.6E-4 1.2E+1 0.0E+0 7.8E-1 1.2E+0 1.4E+03

1967 7.0E-2 1.8E-1 0.0E+0 1.2E-4 1.8E-5 1.7E+0 0.0E+0 2.1E-1 6.6E-1 3.8E+02

1968 5.0E+0 3.4E-1 0.0E+0 2.3E-3 3.4E-4 7.4E-1 0.0E+0 1.2E+0 6.4E-1 2.2E+03

1969 2.8E-1 4.8E-1 0.0E+0 5.0E-4 7.4E-5 3.7E-10.0E+0 3.6E-1 5.3E-1 6.5E+02

1970 6.6E-1 2.5E-5 0.0E+0 7.0E-4 1.1E-4 3.0E-1 0.0E+0 2.7E-1 5.5E-1 4.9E+02

1971 2.5E+0 7.0E-1 0.0E+0 2.1E-3 3.1E-4 3.2E+0 0.0E+0 1.1E+0 4.7E-1 2.0E+03

1972 2.7E-1 2.9E-1 0.0E+0 6.5E-4 9.7E-5 4.9E-1 0.0E+0 2.8E-1 1.7E-1 5.0E+02

1973 5.7E-2 3.2E-5 0.0E+0 5.6E-4 8.1E-5 2.7E-1 0.0E+0 1.5E-1 5.1E-2 2.7E+02

1974 2.4E-2 3.8E-3 0.0E+0 3.3E-4 2.9E-5 1.3E-1 0.0E+0 1.1E-1 3.7E-1 2.0E+02

1975 8.9E-3 4.5E-3 0.0E+0 1.3E-4 2.5E-5 6.4E-2 0.0E+0 1.9E-2 2.5E-1 3.4E+01

1976 7.4E-5 4.6E-5 0.0E+0 1.2E-5 5.1E-6 1.2E-3 0.0E+0 5.5E-4 4.3E-2 9.9E

-01

1977 2.0E-4 1.4E-4 0.0E+0 8.0E-5 3.4E-5 1.1E-2 0.0E+0 5.0E-3 4.3E-1 9.0E+00

1978 3.8E-4 2.0E-3 0.0E+0 7.4E-5 7.9E-6 5.7E-3 0.0E+0 1.9E-3 3.5E-1 3.4E+00

1979 1.8E-4 9.7E-5 0.0E+0 4.8E-5 5.2E-6 1.3E-3 0.0E+0 8.9E-3 5.1E-2 1.6E+01

1980 2.9E-4 1.5E-3 0.0E+0 3.1E-5 4.0E-6 6.3E-4 0.0E+0 4.3E-4 3.1E-1 7.7E

-01

1981 2.9E-4 3.8E-3 0.0E+0 6.1E-6 1.1E-6 6.1E-3 0.0E+0 3.3E-4 2.0E-1 5.9E

-01

1982 1.5E-4 4.7E-5 0.0E+0 1.5E-5 1.6E-6 4.4E-4 0.0E+0 2.8E-4 7.2E-2 5.0

E-01

1983 2.9E-4 1.5E-3 0.0E+0 1.2E-4 1.6E-5 2.3E-3 0.0E+0 1.1E-4 3.6E-2 2.0E

-01

1984 4.1E-4 1.4E-4 0.0E+0 2.7E-5 1.1E-5 4.5E-4 0.0E+0 1.8E-4 2.1E-2 3.2E

-01

1985 4.1E-4 3.0E-3 0.0E+0 7.6E-6 1.5E-6 3.3E-3 0.0E+0 2.2E-4 3.3E-1 4.0E

-01

1986 4.1E-4 1.3E-4 0.0E+0 1.8E-6 1.4E-7 3.3E-3 0.0E+0 2.3E-5 6.2E-2 4.1E

-02

1987 4.1E-4 6.1E-5 0.0E+0 2.0E-6 3.0E-7 4.3E-5 0.0E+0 3.2E-5 1.1E+0 5.8E

-02

1988 1.2E-3 5.9E-5 0.0E+0 4.8E-6 7.3E-7 6.6E-2 0.0E+0 1.2E-4 2.2E+0 2.2E

-01

1989 4.1E-4 1.4E-5 0.0E+0 8.2E-9 1.2E-9 2.3E-4 0.0E+0 1.1E-5 9.6E-2 2.0E

-02
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 23 of 67

Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90Y-91

I-129

1990 4.1E-4 1.3E-4 0.0E+0 3.1E-9 3.1E-9 1.4E-4 0.0E+0 6.6E-7 1.0E-1 1.2E

-03

1991 4.1E-4 5.2E-5 0.0E+0 3.3E-10 3.3E-10 1.7E-4 0.0E+0 3.2E-4 6.0E-2 5.8E

-01

1992 4.1E-4 6.0E-5 0.0E+0 5.7E-7 5.7E-7 4.6E-5 0.0E+0 2.8E-5 1.0E-1 5.0E

-02

1993 0.0E+0 3.8E-5 0.0E+0 0.0E+0 8.7E-10 3.5E-4 0.0E+0 3.8E-4 0.0E+0 6.8E

-01

1994 0.0E+0 1.3E-4 0.0E+0 0.0E+0 2.0E-7 0.0E+0 0.0E+0 3.4E-4 0.0E+0 6.1E

-01

1995 0.0E+0 6.9E-5 0.0E+0 1.1E-7 1.8E-8 0.0E+0 0.0E+0 7.7E-6 0.0E+0 1.4E

-02

1996 0.0E+0 6.5E-5 0.0E+0 5.1E-7 1.0E-8 0.0E+0 0.0E+0 2.5E-6 0.0E+0 4.5E

-03

1997 0.0E+0 0.0E+0 0.0E+0 4.1E-7 1.3E-7 0.0E+0 0.0E+0 5.7E-5 0.0E+0 1.0E

-01

1998 0.0E+0 5.4E-5 0.0E+0 4.0E-7 4.3E-8 0.0E+0 0.0E+0 2.5E-5 0.0E+0 4.5E

-02

1999 0.0E+0 7.2E-5 0.0E+0 1.8E-7 1.7E-8 0.0E+0 0.0E+0 1.0E-5 0.0E+0 1.8E

-02

2000 0.0E+0 6.4E-3 0.0E+0 1.2E-4 1.2E-6 0.0E+0 0.0E+0 1.2E-2 0.0E+0 2.2E+01

2001 0.0E+0 3.3E-3 0.0E+0 8.0E-7 9.5E-5 0.0E+0 0.0E+0 3.9E-4 0.0E+0 7.0E

-01

2002 0.0E+0 5.3E-4 0.0E+0 2.2E-5 6.0E-5 0.0E+0 0.0E+0 1.3E-2 0.0E+0 2.3E+01

2003 1.9E-7 2.4E-2 4.7E-2 2.1E-5 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+00

2004 2.5E-5 1.6E-4 6.7E-4 5.5E-5 7.7E-5 0.0E+0 8.1E-6 1.5E-3 0.0E+0 2.7E+00

2005 8.8E-7 2.3E-2 8.4E-5 4.5E-5 2.6E-4 4.8E-8 3.8E-4 1.3E-2 0.0E+0 2.3E+01

2006–2008 4.96E-6 1.15E-2 1.69E-2 1.90E-5 9.82E-5 1.78E-8 1.44E-4 5.36E-3 0.00E+0 9.6E+00

Table4-6. Intake (Bq/yr) by year for SPERT, 1952 to 2008 [6].

Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90Y-91

I-129

1952 2.4E+1 1.4E-1 0.0E+0 0.0E+0 0.0E+0 1.7E+0 0.0E+0 5.4E-1 2.2E+0 9.7E+02

1953 2.4E+1 3.1E-1 5.7E+0 3.8E-3 5.6E-4 1.7E+0 1.7E+0 2.9E+0 3.7E+0 5.2E+03

1954 6.0E+1 2.3E-1 1.4E+1 9.5E-3 1.4E-3 4.4E+0 4.3E+0 5.1E+0 9.2E+0 9.2E+0

3

1955 8.3E+1 3.4E-1 2.0E+1 9.5E-3 1.4E-3 6.1E+0 6.0E+0 7.3E+0 1.3E+1 1.3E+04

1956 9.6E+1 4.0E+0 2.3E+1 1.5E-2 2.3E-3 7.0E+0 7.2E+0 7.6E+0 1.5E+1 1.4E+04

1957 1.9E+1 4.8E+2 4.8E+1 2.7E-2 4.0E-3 2.0E+0 6.6E+0 1.2E+1 7.3E+0 2.2E+04

1958 2.6E+1 3.6E+2 6.9E+1 3.9E-2 5.7E-3 2.8E+0 5.1E+0 1.8E+1 5.6E+0 3.2E+04

1959 2.1E+1 7.8E+1 5.7E+1 3.1E-2 4.6E-3 2.3E+0 2.6E+0 1.5E+1 2.9E+0 2.7E+04

1960 2.8E-1 1.1E+1 6.0E-1 1.1E-2 1.6E-3 2.8E-2 3.2E-1 9.0E-1 3.5E-1 1.6E+03

1961 2.0E-1 1.5E+1 1.4E-2 2.0E-3 2.9E-4 1.2E-2 8.5E-1 1.1E+0 9.2E-1 2.0E+03

1962 8.9E-1 1.4E+1 2.0E+0 1.3E-3 1.9E-4 8.9E-2 8.6E-1 1.6E+0 9.3E-1 2.9E+03

1963 1.4E+1 9.2E+0 4.0E+1 1.2E-2 1.8E-3 1.6E+0 5.4E-1 1.2E+1 6.0E-1 2.2E+04

1964 7.8E+0 4.8E-1 0.0E+0 4.7E-4 6.9E-5 1.2E+2 1.3E-1 3.0E+0 8.5E+0 5.4E+03

1965 2.0E+1 3.2E+0 0.0E+0 2.0E-2 3.0E-3 8.5E+0 0.0E+0 1.2E+1 7.5E+0 2.2E+04

1966 1.2E+1 1.9E+0 0.0E+0 4.6E-3 6.8E-4 5.3E+1 0.0E+0 3.3E+0 5.0E+0 5.9E+03

1967 3.0E-1 7.9E-1 0.0E+0 5.2E-4 7.6E-5 7.4E+0 0.0E+0 9.0E-1 2.8E+0 1.6E+03

1968 2.1E+1 1.4E+00.0E+0 1.0E-2 1.5E-3 3.2E+0 0.0E+0 5.2E+0 2.7E+0 9.4E+03

1969 1.2E+0 2.1E+0 0.0E+0 2.1E-3 3.2E-4 1.6E+0 0.0E+0 1.5E+0 2.3E+0 2.7E+03

1970 2.8E+0 1.1E-4 0.0E+0 3.0E-3 4.5E-4 1.3E+0 0.0E+0 1.1E+0 2.3E+0 2.0E+03
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 24 of 67

Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90Y-91

I-129

1971 1.1E+1 3.0E+0 0.0E+0 9.0E-3 1.3E-3 1.4E+1 0.0E+0 4.8E+0 2.0E+0 8.6E+03

1972 1.2E+0 1.2E+0 0.0E+0 2.8E-3 4.2E-4 2.1E+0 0.0E+0 1.2E+0 7.2E-1 2.2E+03

1973 8.1E-3 4.6E-6 0.0E+0 8.0E-5 1.2E-5 3.8E-2 0.0E+0 2.1E-2 7.3E-3 3.8E+01

1974 5.5E-2 8.9E-3 0.0E+0 7.8E-4 6.8E-5 3.1E-1 0.0E+0 2.6E-1 8.6E-1 4.7E+02

1975 3.8E-2 1.9E-2 0.0E+0 5.5E-4 1.1E-4 2.7E-1 0.0E+0 8.3E-2 1.1E+0 1.5E+02

1976 7.4E-4 4.6E-4 0.0E+0 1.2E-4 5.1E-5 1.2E-2 0.0E+0 5.5E-3 4.3E-1 9.9E+00

1977 8.7E-4 5.9E-4 0.0E+0 3.4E-4 1.5E-4 4.5E-2 0.0E+0 2.1E-2 1.8E+0 3.8E+01

1978 1.6E-3 8.7E-3 0.0E+0 3.2E-4 3.4E-5 2.4E-2 0.0E+0 8.3E-3 1.5E+0 1.5E+01

1979 2.5E-3 1.4E-3 0.0E+0 6.8E-4 7.4E-5 1.8E-2 0.0E+0 1.3E-1 7.3E-1 2.3E+02

1980 1.2E-3 6.2E-3 0.0E+0 1.3E-4 1.7E-5 2.7E-3 0.0E+0 1.8E-3 1.3E+0 3.2E+00

1981 1.2E-3 1.6E-2 0.0E+0 2.6E-5 4.8E-6 2.6E-2 0.0E+0 1.4E-3 8.6E-1 2.5E+00

1982 2.2E-4 6.7E-5 0.0E+0 2.1E-5 2.3E-6 6.3E-4 0.0E+0 4.0E-4 1.0E-1 7.2E

-01

1983 1.2E-3 6.2E-3 0.0E+0 5.2E-4 6.9E-5 1.0E-2 0.0E+0 4.5E-4 1.5E-1 8.1E

-01

1984 1.2E-3 4.2E-4 0.0E+0 8.0E-5 3.2E-5 1.4E-3 0.0E+0 5.5E-4 6.2E-2 9.9E

-01

1985 1.2E-3 9.0E-3 0.0E+0 2.3E-5 4.5E-6 1.0E-2 0.0E+0 6.6E-4 9.9E-1 1.2E+00

1986 1.2E-3 3.8E-4 0.0E+0 5.5E-6 4.2E-7 1.0E-2 0.0E+0 6.9E-5 1.9E-1 1.2E

-01

1987 1.2E-3 1.8E-4 0.0E+0 5.9E-6 9.0E-7 1.3E-4 0.0E+0 9.7E-5 3.3E+0 1.7E

-01

1988 4.1E-4 2.0E-5 0.0E+0 1.6E-6 2.4E-7 2.2E-2 0.0E+0 3.9E-5 7.4E-1 7.0E

-02

1989 1.2E-3 4.2E-5 0.0E+0 2.5E-8 3.5E-9 6.9E-4 0.0E+0 3.2E-5 2.9E-1 5.8E

-02

1990 4.1E-4 1.3E-4 0.0E+0 3.1E-9 3.1E-9 1.4E-4 0.0E+0 6.6E-7 1.0E-1 1.2E

-03

1991 4.1E-4 5.2E-5 0.0E+0 3.3E-10 3.3E-10 1.7E-4 0.0E+0 3.2E-4 6.0E-2 5.8E

-01

1992 1.2E-3 1.8E-4 0.0E+0 1.7E-6 1.7E-6 1.4E-4 0.0E+0 8.3E-5 3.1E-1 1.5E

-01

1993 0.0E+0 3.8E-5 0.0E+0 0.0E+0 8.7E-10 3.5E-4 0.0E+0 3.8E-4 0.0E+0 6.8E

-01

1994 0.0E+0 1.3E-4 0.0E+0 0.0E+0 2.0E-7 0.0E+0 0.0E+0 3.4E-4 0.0E+0 6.1E

-01

1995 0.0E+0 6.9E-5 0.0E+0 1.1E-7 1.8E-8 0.0E+0 0.0E+0 7.7E-6 0.0E+0 1.4E

-02

1996 0.0E+0 6.5E-5 0.0E+0 5.1E-7 1.0E-8 0.0E+0 0.0E+0 2.5E-6 0.0E+0 4.5E

-03

1997 0.0E+0 0.0E+0 0.0E+0 4.1E-7 1.3E-7 0.0E+0 0.0E+0 5.7E-5 0.0E+0 1.0E

-01

1998 0.0E+0 5.4E-5 0.0E+0 4.0E-7 4.3E-8 0.0E+0 0.0E+0 2.5E-5 0.0E+0 4.5E

-02

1999 0.0E+0 3.1E-4 0.0E+0 7.5E-7 7.1E-8 0.0E+0 0.0E+0 4.4E-5 0.0E+0 7.9E

-02

2000 0.0E+0 6.4E-3 0.0E+0 1.2E-4 1.2E-6 0.0E+0 0.0E+0 1.2E-2 0.0E+0 2.2E+01

2001 0.0E+0 3.3E-3 0.0E+0 8.0E-7 9.5E-5 0.0E+0 0.0E+0 3.9E-4 0.0E+0 7.0E

-01

2002 0.0E+0 3.7E-4 0.0E+0 1.6E-5 4.2E-5 0.0E+0 0.0E+0 8.9E-3 0.0E+0 1.6E+01

2003 1.1E-7 1.4E-2 2.8E-2 1.3E-5 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+00

2004 1.1E-5 7.4E-5 3.0E-4 2.5E-5 3.5E-5 0.0E+0 3.7E-6 6.7E-4 0.0E+0 1.2E+00

2005 6.3E-7 1.6E-2 6.0E-5 3.2E-5 1.9E-4 3.5E-8 2.7E-4 9.6E-3 0.0E+0 1.7E+01

2006–2008 1.95E-6 3.52E-3 4.98E-3 2.09E-5 3.55E-5 6.24E-9 4.80E-5 2.80E-3 0.00E+0 5.0E+00
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 25 of 67

Table4-7. Intake (Bq/yr) by year for TAN, 1952 to 2008 [6].

Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90Y-91

I-129

1952 5.5E+0 3.3E-2 0.0E+0 0.0E+0 0.0E+0 4.1E-1 0.0E+0 1.3E-1 5.0E-1 2.3E+02

1953 5.5E+0 7.3E-2 1.3E+0 8.9E-4 1.3E-4 4.1E-14.0E-1 6.8E-1 8.6E-1 1.2E+03

1954 1.4E+1 5.3E-2 3.3E+0 2.2E-3 3.3E-4 1.0E+0 9.9E-1 1.2E+0 2.1E+0 2.2E+03

1955 1.9E+1 7.9E-2 4.7E+0 2.2E-3 3.3E-4 1.4E+0 1.4E+0 1.7E+0 3.0E+0 3.1E+03

1956 2.2E+1 9.4E-1 5.4E+0 3.6E-3 5.3E-4 1.6E+0 1.7E+0 1.8E+0 3.5E+0 3.2E+03

1957 4.4E+0 1.1E+2 1.1E+1 6.3E-3 9.3E-4 4.7E-1 1.6E+0 2.9E+0 1.7E+0 5.2E+03

1958 6.0E+0 8.3E+1 1.6E+1 9.1E-3 1.3E-3 6.6E-1 1.2E+0 4.1E+0 1.3E+0 7.4E+03

1959 4.8E+0 1.8E+1 1.3E+1 7.3E-3 1.1E-3 5.4E-1 6.1E-1 3.4E+0 6.7E-1 6.1E+03

1960 6.6E-2 2.6E+0 1.4E-1 2.5E-3 3.7E-4 6.6E-3 7.6E-2 2.1E-1 8.2E-2 3.8E+02

1961 4.6E-2 3.5E+0 3.3E-3 4.6E-4 6.8E-5 2.8E-3 2.0E-1 2.5E-1 2.1E-1 4.5E+02

1962 2.1E-1 3.3E+0 4.6E-1 3.0E-4 4.5E-5 2.1E-2 2.0E-1 3.8E-1 2.2E-1 6.8E+02

1963 3.3E+0 2.1E+0 9.4E+0 2.8E-3 4.2E-4 3.7E-1 1.3E-1 2.8E+0 1.4E-1 5.0E+03

1964 1.8E+0 1.1E-1 0.0E+0 1.1E-4 1.6E-5 2.8E+1 3.0E-2 7.1E-1 2.0E+0 1.3E+03

1965 4.6E+0 7.4E-1 0.0E+0 4.7E-3 7.0E-4 2.0E+0 0.0E+0 2.7E+0 1.8E+0 4.9E+03

1966 2.8E+0 4.3E-1 0.0E+0 1.1E-3 1.6E-4 1.2E+1 0.0E+0 7.8E-1 1.2E+0 1.4E+03

1967 7.0E-2 1.8E-1 0.0E+0 1.2E-4 1.8E-5 1.7E+0 0.0E+0 2.1E-1 6.6E-1 3.8E+02

1968 5.0E+0 3.4E-1 0.0E+0 2.3E-3 3.4E-4 7.4E-1 0.0E+0 1.2E+0 6.4E-1 2.2E+03

1969 2.8E-1 4.8E-1 0.0E+0 5.0E-4 7.4E-5 3.7E-1 0.0E+0 3.6E-1 5.3E-1 6.5E+02

1970 6.6E-1 2.5E-5 0.0E+0 7.0E-4 1.1E-4 3.0E-1 0.0E+0 2.7E-1 5.5E-1 4.9E+02

1971 2.5E+0 7.0E-1 0.0E+0 2.1E-3 3.1E-4 3.2E+0 0.0E+0 1.1E+0 4.7E-1 2.0E+03

1972 2.7E-1 2.9E-1 0.0E+0 6.5E-4 9.7E-5 4.9E-1 0.0E+0 2.8E-1 1.7E-1 5.0E+02

1973 2.4E-2 1.4E-5 0.0E+0 2.4E-4 3.5E-5 1.1E-1 0.0E+0 6.2E-2 2.2E-2 1.1E+02

1974 7.9E-3 1.3E-3 0.0E+0 1.1E-4 9.7E-6 4.4E-2 0.0E+0 3.7E-2 1.2E-1 6.7E+01

1975 3.8E-3 1.9E-3 0.0E+0 5.5E-5 1.1E-5 2.7E-2 0.0E+0 8.3E-3 1.1E-1 1.5E+01

1976 7.4E-5 4.6E-5 0.0E+0 1.2E-5 5.1E-6 1.2E-3 0.0E+0 5.5E-4 4.3E-2

9.9E-01

1977 2.9E-5 2.0E-5 0.0E+0 1.1E-5 4.8E-6 1.5E-3 0.0E+0 7.2E-4 6.1E-2 1.3E+00

1978 3.8E-4 2.0E-3 0.0E+0 7.4E-5 7.9E-6 5.7E-3 0.0E+0 1.9E-3 3.5E-1 3.4E+00

1979 7.6E-5 4.2E-5 0.0E+0 2.0E-5 2.2E-6 5.5E-4 0.0E+0 3.8E-3 2.2E-2 6.8E+00

1980 1.2E-4 6.2E-4 0.0E+0 1.3E-5 1.7E-6 2.7E-4 0.0E+0 1.8E-4 1.3E-1 3.2E

-01

1981 1.2E-4 1.6E-3 0.0E+0 2.6E-6 4.8E-7 2.6E-3 0.0E+0 1.4E-4 8.6E-2 2.5E

-01

1982 6.6E-5 2.0E-5 0.0E+0 6.2E-6 6.9E-7 1.9E-4 0.0E+0 1.2E-4 3.1E-2 2.2E

-01

1983 1.2E-4 6.2E-4 0.0E+0 5.2E-5 6.9E-6 1.0E-3 0.0E+0 4.5E-5 1.5E-2 8.1E

-02

1984 1.2E-4 4.2E-5 0.0E+0 8.0E-6 3.2E-6 1.4E-4 0.0E+0 5.5E-5 6.2E-3 9.9E

-02

1985 1.2E-4 9.0E-4 0.0E+0 2.3E-6 4.5E-7 1.0E-3 0.0E+0 6.6E-5 9.9E-2 1.2E

-01

1986 1.2E-4 3.8E-5 0.0E+0 5.5E-7 4.2E-8 1.0E-3 0.0E+0 6.9E-6 1.9E-2 1.2E

-02

1987 2.9E-4 4.3E-5 0.0E+0 1.4E-6 2.1E-7 3.0E-5 0.0E+0 2.3E-5 7.6E-1 4.1E

-02

1988 1.2E-4 5.9E-6 0.0E+0 4.8E-7 7.3E-8 6.6E-3 0.0E+0 1.2E-5 2.2E-1 2.2E

-02

1989 2.9E-4 9.7E-6 0.0E+0 5.7E-9 8.1E-10 1.6E-4 0.0E+0 7.4E-6 6.7E-2 1.3E

-02
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Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90Y-91

I-129

1990 1.2E-4 3.8E-5 0.0E+0 9.4E-10 9.4E-10 4.2E-5 0.0E+0 2.0E-7 3.1E-2 3.6E

-04

1991 1.2E-4 1.6E-5 0.0E+0 1.0E-10 1.0E-10 5.2E-5 0.0E+0 9.7E-5 1.8E-2 1.7E

-01

1992 1.2E-4 1.8E-5 0.0E+0 1.7E-7 1.7E-7 1.4E-5 0.0E+0 8.3E-6 3.1E-2 1.5E

-02

1993 0.0E+0 3.8E-6 0.0E+0 0.0E+0 8.7E-11 3.5E-5 0.0E+0 3.8E-5 0.0E+0 6.8E

-02

1994 0.0E+0 3.1E-5 0.0E+0 0.0E+0 4.6E-8 0.0E+0 0.0E+0 8.0E-5 0.0E+0 1.4E

-01

1995 0.0E+0 4.8E-5 0.0E+0 7.7E-8 1.3E-8 0.0E+0 0.0E+0 5.4E-6 0.0E+0 9.7E

-03

1996 0.0E+0 2.8E-5 0.0E+0 2.2E-7 4.5E-9 0.0E+0 0.0E+0 1.1E-6 0.0E+0 2.0E

-03

1997 0.0E+0 0.0E+0 0.0E+0 1.8E-7 5.5E-8 0.0E+0 0.0E+0 2.4E-5 0.0E+0 4.3E

-02

1998 0.0E+0 2.3E-5 0.0E+0 1.7E-7 1.8E-8 0.0E+0 0.0E+0 1.1E-5 0.0E+0 2.0E

-02

1999 0.0E+0 3.1E-5 0.0E+0 7.5E-8 7.1E-9 0.0E+0 0.0E+0 4.4E-6 0.0E+0 7.9E

-03

2000 0.0E+0 1.9E-2 0.0E+0 3.6E-4 3.6E-6 0.0E+0 0.0E+0 3.5E-2 0.0E+0 6.3E+01

2001 0.0E+0 1.0E-2 0.0E+0 2.4E-6 2.9E-4 0.0E+0 0.0E+0 1.2E-3 0.0E+0 2.2E+00

2002 0.0E+0 1.6E-4 0.0E+0 6.6E-6 1.8E-5 0.0E+0 0.0E+0 3.8E-3 0.0E+0 6.8E+00

2003 3.8E-8 4.8E-3 9.5E-3 4.2E-6 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+00

2004 3.4E-6 2.2E-5 9.1E-5 7.6E-6 1.0E-5 0.0E+0 1.1E-6 2.0E-4 0.0E+0 3.6E-01

2005 1.9E-7 4.9E-3 1.8E-5 9.6E-6 5.6E-5 1.0E-8 8.1E-5 2.9E-3 0.0E+0 5.2E+00

2006–2008 6.03E-7 3.59E-3 1.69E-3 6.36E-5 5.11E-5 1.78E-9 1.44E-5 6.14E-3 0.00E+0 1.1E+01

Table4-8. Intake (Bq/yr) by year for TRA, 1952 to 2008 [6].

Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90Y-91

I-129

1952 2.4E+1 1.4E-1 0.0E+0 0.0E+0 0.0E+0 1.7E+0 0.0E+0 5.4E-1 2.2E+0 9.7E+02

1953 2.4E+1 3.1E-1 5.7E+0 3.8E-3 5.6E-4 1.7E+0 1.7E+0 2.9E+0 3.7E+0 5.2E+03

1954 6.0E+1 2.3E-1 1.4E+1 9.5E-3 1.4E-3 4.4E+0 4.3E+0 5.1E+0 9.2E+0 9.2E+03

1955 8.3E+1 3.4E-1 2.0E+1 9.5E-3 1.4E-3 6.1E+0 6.0E+0 7.3E+0 1.3E+1 1.3E+04

1956 9.6E+1 4.0E+0 2.3E+1 1.5E-2 2.3E-3 7.0E+0 7.2E+0 7.6E+0 1.5E+1 1.4E+04

1957 1.9E+1 4.8E+2 4.8E+1 2.7E-2 4.0E-3 2.0E+0 6.6E+0 1.2E+1 7.3E+0 2.2E+04

1958 2.6E+1 3.6E+2 6.9E+1 3.9E-2 5.7E-3 2.8E+0 5.1E+0 1.8E+1 5.6E+0 3.2E+04

1959 2.1E+1 7.8E+1 5.7E+1 3.1E-2 4.6E-3 2.3E+0 2.6E+0 1.5E+1 2.9E+0 2.7E+04

1960 2.8E-1 1.1E+1 6.0E-1 1.1E-2 1.6E-3 2.8E-2 3.2E-1 9.0E-1 3.5E-1 1.6E+03

1961 2.0E-1 1.5E+1 1.4E-2 2.0E-3 2.9E-4 1.2E-2 8.5E-1 1.1E+0 9.2E-1 2.0E+03

1962 8.9E-1 1.4E+1 2.0E+0 1.3E-3 1.9E-4 8.9E-2 8.6E-1 1.6E+0 9.3E-1 2.9E+03

1963 1.4E+1 9.2E+0 4.0E+1 1.2E-2 1.8E-3 1.6E+0 5.4E-1 1.2E+1 6.0E-1 2.2E+04

1964 7.8E+0 4.8E-1 0.0E+0 4.7E-4 6.9E-5 1.2E+2 1.3E-1 3.0E+0 8.5E+0 5.4E+03

1965 2.0E+1 3.2E+0 0.0E+0 2.0E-2 3.0E-3 8.5E+0 0.0E+0 1.2E+1 7.5E+0 2.2E+04

1966 1.2E+1 1.9E+0 0.0E+0 4.6E-3 6.8E-4 5.3E+1 0.0E+0 3.3E+0 5.0E+0 5.9E+03

1967 3.0E-1 7.9E-1 0.0E+0 5.2E-4 7.6E-5 7.4E+0 0.0E+0 9.0E-1 2.8E+0 1.6E+03

1968 2.1E+1 1.4E+0 0.0E+0 1.0E-2 1.5E-3 3.2E+0 0.0E+0 5.2E+0 2.7E+0

9.4E+03

1969 1.2E+0 2.1E+0 0.0E+0 2.1E-3 3.2E-4 1.6E+0 0.0E+0 1.5E+0 2.3E+0 2.7E+03

1970 2.8E+0 1.1E-4 0.0E+0 3.0E-3 4.5E-4 1.3E+0 0.0E+0 1.1E+0 2.3E+0 2.0E+03
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Nuclide Ce-144I-131 Pm-147 Pu-238 Pu-239, 240 Ru-106 Sr-89 Sr-90Y-91

I-129

1971 1.1E+1 3.0E+0 0.0E+0 9.0E-3 1.3E-3 1.4E+1 0.0E+0 4.8E+0 2.0E+0 8.6E+03

1972 1.2E+0 1.2E+0 0.0E+0 2.8E-3 4.2E-4 2.1E+0 0.0E+0 1.2E+0 7.2E-1 2.2E+03

1973 5.7E-2 3.2E-5 0.0E+0 5.6E-4 8.1E-5 2.7E-1 0.0E+0 1.5E-1 5.1E-2 2.7E+02

1974 2.4E-2 3.8E-3 0.0E+0 3.3E-4 2.9E-5 1.3E-1 0.0E+0 1.1E-1 3.7E-1 2.0E+02

1975 8.9E-3 4.5E-3 0.0E+0 1.3E-4 2.5E-5 6.4E-2 0.0E+0 1.9E-2 2.5E-1 3.4E+01

1976 2.2E-4 1.4E-4 0.0E+0 3.5E-5 1.5E-5 3.5E-3 0.0E+0 1.7E-3 1.3E-1 3.1E+00

1977 2.9E-4 2.0E-4 0.0E+0 1.1E-4 4.8E-5 1.5E-2 0.0E+0 7.2E-3 6.1E-1 1.3E+01

1978 1.6E-3 8.7E-3 0.0E+0 3.2E-4 3.4E-5 2.4E-2 0.0E+0 8.3E-3 1.5E+0 1.5E+01

1979 1.8E-4 9.7E-5 0.0E+0 4.8E-5 5.2E-6 1.3E-3 0.0E+0 8.9E-3 5.1E-2 1.6E+01

1980 2.9E-4 1.5E-3 0.0E+0 3.1E-5 4.0E-6 6.3E-4 0.0E+0 4.3E-4 3.1E-1 7.7E

-01

1981 2.9E-4 3.8E-3 0.0E+0 6.1E-6 1.1E-6 6.1E-3 0.0E+0 3.3E-4 2.0E-1 5.9E

-01

1982 1.5E-4 4.7E-5 0.0E+0 1.5E-5 1.6E-6 4.4E-4 0.0E+0 2.8E-4 7.2E-2 5.0E

-01

1983 2.9E-4 1.5E-3 0.0E+0 1.2E-4 1.6E-5 2.3E-3 0.0E+0 1.1E-4 3.6E-2 2.0E

-01

1984 2.9E-4 9.7E-5 0.0E+0 1.9E-5 7.4E-6 3.2E-4 0.0E+0 1.3E-4 1.4E-2 2.3E

-01

1985 4.1E-4 3.0E-3 0.0E+0 7.6E-6 1.5E-6 3.3E-3 0.0E+0 2.2E-4 3.3E-1 4.0E

-01

1986 4.1E-4 1.3E-4 0.0E+0 1.8E-6 1.4E-7 3.3E-3 0.0E+0 2.3E-5 6.2E-2 4.1E

-02

1987 1.2E-3 1.8E-4 0.0E+0 5.9E-6 9.0E-7 1.3E-4 0.0E+0 9.7E-5 3.3E+0 1.7E

-01

1988 1.2E-3 5.9E-5 0.0E+0 4.8E-6 7.3E-7 6.6E-2 0.0E+0 1.2E-4 2.2E+0 2.2E

-01

1989 1.2E-3 4.2E-5 0.0E+0 2.5E-8 3.5E-9 6.9E-4 0.0E+0 3.2E-5 2.9E-1 5.8E

-02

1990 4.1E-4 1.3E-4 0.0E+0 3.1E-9 3.1E-9 1.4E-4 0.0E+0 6.6E-7 1.0E-1 1.2E

-03

1991 4.1E-4 5.2E-5 0.0E+0 3.3E-10 3.3E-10 1.7E-4 0.0E+0 3.2E-4 6.0E-2 5.8E

-01

1992 4.1E-4 6.0E-5 0.0E+0 5.7E-7 5.7E-7 4.6E-5 0.0E+0 2.8E-5 1.0E-1 5.0E

-02

1993 0.0E+0 3.8E-5 0.0E+0 0.0E+0 8.7E-10 3.5E-4 0.0E+0 3.8E-4 0.0E+0 6.8E

-01

1994 0.0E+0 4.4E-5 0.0E+0 0.0E+0 6.6E-8 0.0E+0 0.0E+0 1.1E-4 0.0E+0 2.0E

-01

1995 0.0E+0 4.8E-5 0.0E+0 7.7E-8 1.3E-8 0.0E+0 0.0E+0 5.4E-6 0.0E+0 9.7E

-03

1996 0.0E+0 6.5E-5 0.0E+0 5.1E-7 1.0E-8 0.0E+0 0.0E+0 2.5E-6 0.0E+0 4.5E

-03

1997 0.0E+0 0.0E+0 0.0E+0 4.1E-7 1.3E-7 0.0E+0 0.0E+0 5.7E-5 0.0E+0 1.0E

-01

1998 0.0E+0 5.4E-5 0.0E+0 4.0E-7 4.3E-8 0.0E+0 0.0E+0 2.5E-5 0.0E+0 4.5E

-02

1999 0.0E+0 7.2E-5 0.0E+0 1.8E-7 1.7E-8 0.0E+0 0.0E+0 1.0E-5 0.0E+0 1.8E

-02

2000 0.0E+0 6.4E-3 0.0E+0 1.2E-4 1.2E-6 0.0E+0 0.0E+0 1.2E-2 0.0E+0 2.2E+01

2001 0.0E+0 3.3E-3 0.0E+0 8.0E-7 9.5E-5 0.0E+0 0.0E+0 3.9E-4 0.0E+0 7.0E

-01

2002 0.0E+0 1.6E-3 0.0E+0 6.7E-5 1.8E-4 0.0E+0 0.0E+0 3.8E-2 0.0E+0 6.8E+01

2003 2.8E-7 3.6E-2 7.1E-2 3.1E-5 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+0 0.0E+00

2004 3.4E-5 2.2E-4 9.1E-4 7.6E-5 1.0E-4 0.0E+0 1.1E-5 2.0E-3 0.0E+0 3.6E+00

2005 1.9E-6 4.9E-2 1.8E-4 9.6E-5 5.6E-4 1.0E-7 8.1E-4 2.9E-2 0.0E+0 5.2E+01

2006–2008 6.03E-6 1.01E-2 1.26E-2 2.70E-5 9.98E-5 1.78E-8 1.44E-4 7.99E-3 0.00E+0 1.4E+01
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Environmental air sampling at the facility areas has been performed at least since the mid-1950s 

where airborne effluents were known or suspected to exist [7]. The early IDO Health and Safety 

(H&S) Division EMRs (AEC 1960, 1961a,b, 1962a,b, 1963, 1964a,b, 1965a,b, 1966a,b, 1967a,b, 

1968a,b, 1969a,b, 1970a,b) document many studies for defining radionuclide concentrations in the 

vicinity of different facilities. These studies were specific for a given test, operation, or incident,

however, and were not performed in a set facility location or for a standard duration. Some facility 

data are presented in the 1963 Annual Progress Report of the H&S Division (Dodd 1964) and a 

routine facility environmental monitoring program was developed between 1963 and 1970. In 1968 

and 1969, formal annual EMRs reported alpha, beta, and 131I concentrations that can be correlated 

with the values in Tables 4-1 through 4-8. The 1970 EMR (AEC 1971, p. 5) discusses gross beta 

values measured at CFA that can be correlated with Table 4-3 (CFA) values. EMRs between 1970 

and 1990 were reviewed for data that could be used for this correlation (AEC 1971, 1972, 1973, 1974; 

ERDA 1975, 1976, 1977b; DOE 1978a,b, 1979a, 1980a, 1981a, 1982a, 1983a; Hoff, Chew, and 

Dickson 1984, 1985; Hoff, Chew, and Rope 1986, 1987; Chew and Mitchell 1988; Hoff, Mitchell, and 

Moore 1989; Hoff et al. 1990, 1991).

Table 4-9 lists results of the comparison. Because of the large variation in measurements made, the 

ratio of values calculated from the EMRs (column 5) to that derived from releases listed in Table 4-1 

through 4-8 is not well-behaved. Three of the 13 values with uncertainties overlap. The geometric 

mean of the remainder of values is 0.36 with a geometric standard deviation (GSD) of 5.3. 

Nevertheless, this comparison provides confidence in the results of Table 4-1 through 4-8. The value 

at the upper 84% confidence value, 1.9 (0.36 × 5.3), is less than the default GSD of 3 assumed for 

environmental results [8].

Figure 4-4 shows the variation of the INL environmental monitoring sampling results for the 9-year

period from 1978 through 1986. This figure also shows the close correlation of environmental sample 

results acquired at distant communities and those acquired at INL facilities as well as the effect of 

foreign nuclear tests and the Chernobyl reactor accident on INL environmental sampling results. As 

shown on this figure, the INL average concentration for the 9-year period has not differed from the 

distant community concentrations by more than a factor of 2. Inspection of subsequent EMRs shows 

the same is true for the years after 1986 (Hoff, Chew, and Rope 1987; Chew and Mitchell 1988; Hoff, 

Mitchell, and Moore 1989; Hoff et al. 1990, 1991, 1992, 1993; Mitchell 1994; Mitchell, Peterson, and 

Hoff 1995; Mitchell et al. 1996, 1997; Evans et al. 1998; Saffel et al. 2000; Stoller 2002a,b,c, 2003).

It is also interesting to note that the greater perturbations in the facility and distant community 

concentrations nearly all correlate with fallout from nuclear tests. Figure 4-5 (Williamson 1977a, p. 

13) shows an example of such a perturbation, attributed to a September 26, 1976, atmospheric test 

conducted by the People’s Republic of China, where the normal concentration was increased by a 

factor of 20 and had a 3-month influence on the average concentration for all the air concentrations 

measured, on and off the site. In the history of air monitoring at INL, an operational release has never 

approached the magnitude of perturbation created by fallout.

4.2.2

4.2.2.1 Stationary Low-Power Reactor No. 1 Accident

Episodic Releases at INL

One significant accident at INL in the last 51 years released substantial amounts of radioactive 

material to the environment. On January 3, 1961, a steam explosion at the SL-1 facility (near the 

present location of ARA II in Figure 4-1) killed three personnel and ruptured the SL-1 reactor vessel. 

The rupture propelled radioactive material into the reactor building and then into the environment. 

The amount of the release and the path the cloud traveled from the reactor building were carefully 

monitored and well-documented (Gammill 1961; Horan and Gammill 1961; Kunze 1962). All 
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 29 of 67

Table 4-9. Comparison of calculated facility intakes with intakes from environmental monitoring 

results. 

Year Activity type

Average 

annual

concentration Reference for Col. 3

Annual 

inhaled

quantity 

(Bq)

INL TBD 

table inhaled 

quantity (Bq)

1963 β-γ 17 pCi/m3 1,510 1,310a

Pu-239 0.6 fCi/m3 0.05 0.014a

1968 α 2.2 fCi/m3 AEC 1969a, p. 14 0.18 0.01a

β 0.64 pCi/m3 AEC 1969a, p. 14 56 337a

I-131 <0.08 pCi/m3 AEC 1969a, p. 14 <7.1 1.4a

1969 α 0.003 pCi/m3 AEC 1970a, p. 14 0.27 2.4E-3a

β 0.501 pCi/m3 AEC 1970a, p. 14 44 118a

I-131 <0.014 pCi/m3 AEC 1970a, p. 14 <1.2 2.1a

1970 Gross β 0.6 pCi/m3 AEC 1971, p. 5 53 74a

Max. gr. β at CFA 0.81 pCi/m3 AEC 1971, p. 5 72 74a

1973 Gross β 95 + 42 fCi/m AEC 1974, p. 14 3 8.4 ± 3.7 0.8a

EBR-I Sr-90 3.4 ±3.0 fCi/m3 AEC 1974, p. 16 0.3 ± 0.27 0.15b

Nb-95 1.0-2.5 fCi/m3 AEC 1974, p. 16 0.09–0.22 ---

Cs-137 7-17 fCi/m3 AEC 1974, p. 16 0.6–1.5 ---

Ce-144 4-8 fCi/m3 AEC 1974, p. 16 0.36–0.71 0.057b

EFS Sr-90 5.9±8.6 fCi/m3 AEC 1974, p. 16 0.52 ± 0.76 0.15a

Nb-95 0.9-2.4 fCi/m3 AEC 1974, p. 16 0.08–0.21 ---

Ru-106 6-9.8 fCi/m3 AEC 1974, p. 16 0.53–0.87 0.27a

Cs-134 0.8-1.6 fCi/m3 AEC 1974, p. 16 0.07–0.14 ---

Cs-137 17-27 fCi/m3 AEC 1974, p. 16 1.5–2.4 ---

1976c Gross β 30-60 fCi/m3 ERDA 1977b 2.5–5.3 0.6-25d

1986 Kr-85 at CFA 37±15 pCi/m3 Hoff, Chew, and Rope 1987 3,290 ± 1330 890e

1988 Kr-85 at CFA 108 ± 69 pCi/m3 Hoff, Chew, and Moore 1989 9,770 ± 6130 14,000e

1990 Kr-85 at CFA 27.7 pCi/m3 Hoff et al. 1992 2,400 690e

a. Values from INEEL TBD Table 4-3 for CFA.

b. Values from INEEL TBD Table 4-5 for RWMC since EBR-I is near RWMC.

c. Of 90 monthly values (January through September) for 10 facility areas, 89 values were between 3 × 10-14 and 6 × 10-14

μCi/mL.

d. Using the current tables with 11 radionuclides, the inhaled quantity is about 0.6 Bq; with the original tables with 44 

radionuclides, the inhaled quantity is about 25 Bq.

e. Values derived from tables in Peterson (2004) that contained concentrations of all INEEL released radionuclides.

radiological doses to personnel involved in the rescue and the cleanup of the reactor building were 

carefully controlled and documented. The SL-1 accident did not affect any other INL facility with the 

effluent of radioactive material. The effluent traveled to the south of the facility, as shown in 

Figure 4-6 (DOE 1991).

4.2.2.2 Criticalities

Three accidental criticalities have occurred at INTEC. The first occurred on October 16, 1959; the 

second on January 25, 1961; and the third on October 17, 1978. The 1978 criticality released 

essentially just the noble fission gases produced during the criticality; halogens and solids were 

removed by the exhaust filtering system (Casto 1980). However, the two earlier criticalities did 

release radioactive material during or shortly after the event; in both cases, the effluent was 

transported to the south-southwest and potentially exposed personnel at the south end of INTEC and 

at CFA and at the EBR-I facility, because that facility was postulated to be in the plume path. A 

conservative analysis, described below, defined the potential radiological exposures that could have 

occurred to individuals at these locations. An additional critical excursion occurred in 1958 as part of 

the critical experiments and low-power testing phase of the HTRE No. 3 reactor configuration for the 

GE-ANP Program. 
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 30 of 67

Figure 4-4. Onsite and distant particulate beta concentrations in air (Hoff, Chew, and 

Rope 1987, p. 17).

4.2.2.2.1 INTEC Criticality of October 16, 1959

On October 16, 1959, at approximately 3:00 a.m., a criticality event occurred in the WH-100 vessel. 

The estimated magnitude of this event was no greater than 4 × 1019 fissions (DOE 1991). Nuclear 

Incident at the Idaho Chemical Processing Plant (Ginkel et al. 1960) gives a full account of the 

incident and documents the radiological doses (calculated internal and measured external) for plant 

personnel involved in the incident. For the calculation of intakes for this incident, meteorological 
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 31 of 67

Figure 4-5. CFA air-monitoring (gross beta) data for 1976 (Williamson 1977a, p. 13).

conditions were modeled so that the χ/Q at 22 km matched the value calculated for Frenchman’s 

Cabin (south of INL, as shown on Figure 4-1), where offsite doses were calculated and reported in 

DOE (1991). The RSAC-6 computer program was used to calculate χ/Q values for the south end of 

INTEC and CFA (Peterson 2004). These concentrations and intake quantities would be applicable 

only if the individual was in the respective areas on the morning of October 16, 1959. Table 4-10 lists 

the intakes applicable at the south end of INTEC, CFA, and EBR-1. 
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Figure 4-6. Dispersion coefficient contours for the SL-1 accident (redrafted from Horan and 

Gammill 1961).

4.2.2.2.2 INTEC Criticality of January 25, 1961

The January 25, 1961, criticality occurred in vessel H-110 about 9:50 a.m. This event consisted of an 

estimated 6.0 × 1017 fissions. The report documenting the incident stated:

Of the 251 individuals present in the ICPP area at the time of the incident, none 

received significant radiation exposure. The highest exposure as determined from film 

badge readings did not exceed 55 millirem of penetrating radiation. Essentially no beta 

radiation was detected. No significant neutron exposure or internal contamination from 

inhalation was found. The absence of significant exposures is attributable to the 

extensive shielding provided by the process cell in which the event took place and the 

control of the fission gases by the equipment. (Paulus et al. 1961)

As for the 1959 criticality, Χ/Q values were calculated for the south end of INTEC, CFA, and EBR-1. 

The source term used for this event is the same as that used for DOE (1991). Table 4-10 lists the 

intakes applicable at the south end of INTEC, CFA, and EBR-1 if the individual was in the respective 

areas on January 25, 1961. 

4.2.2.2.3 INTEC Criticality of October 17, 1978

At approximately 8:40 p.m. on October 17, 1978, a criticality incident occurred in the first-cycle 

uranium extraction system in the CPP-601 process building (DOE 1979a; Casto 1980). The fissioning 

lasted for about 20 minutes and involved about 3.0 × 1018 fissions. This event did not result in 

significant radiation exposures to personnel, and there was no contamination of general plant areas. 
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Table 4-10. Intakes (Bq/event) for criticalities that occurred at INL [9].

Date 11/18/1958 10/16/1959 10/16/1959 10/16/1959 1/25/1961 1/25/1961 1/25/1961 10/17/1978 10/17/1978

Event IET 13 Criticality Criticality Criticality Criticality Criticality Criticality Criticality Criticality

Exposure location LPTF/STPF South ICPP CFA area EBR-I area South ICPP CFA area EBR-I area South ICPP CFA area

Rb-89 8.1E+2 5.2E+5 8.8E+4 2.1E+4 1.5E+1 1.0E+0 4.4E-1 2.2E+3 3.7E+4

Sr-91 4.1E+2 5.4E+4 9.2E+3 2.2E+3 1.7E+2 1.1E+1 4.8E+0 1.7E+1 7.0E+2

Sr-92 1.9E+2 6.5E+4 1.1E+4 2.6E+3 1.5E+2 9.7E+0 4.3E+0 5.6E+0 2.1E+2

Y-92 5.6E+1 7.5E+3 1.3E+3 3.1E+2 7.3E+1 4.9E+0 2.2E+0 9.1E-2 2.6E+1

Y-93 2.0E+1 5.9E+3 1.0E+3 2.4E+2 1.9E+1 1.3E+0 5.7E-1 1.3E-1 1.5E+1

Te-133 1.1E+3 8.0E+3

I-131(elem.) 3.4E+0 2.9E+2 5.0E+1 1.2E+1 1.9E+0 1.3E-1 5.7E-2 3.9E-2 1.7E+0

I-133 7.1E+1 6.2E+3 1.1E+3 2.6E+2 3.8E+1 2.5E+0 1.1E+0 4.4E+0 5.8E+2

I-134 2.0E+2 4.6E+4 7.8E+3 1.9E+3 7.2E+1 4.8E+0 2.1E+0 1.1E+2 3.4E+3

I-135 4.6E+2 2.0E+4 3.4E+3 8.2E+2 1.1E+2 7.2E+0 3.2E+0 5.5E+1 2.2E+3

Cs-138 5.0E+3 7.7E+5 1.3E+5 3.1E+4 5.2E+2 5.6E+4

Ba-139 3.4E+3 4.0E+5 6.8E+4 1.6E+4 6.6E+2 4.4E+1 1.9E+1 1.4E+2 1.8E+4

La-141 2.0E+2 2.8E+4 4.7E+3 1.1E+3 1.0E+2 6.8E+0 3.0E+0 4.6E-1 9.0E+1

La-142 9.3E+1 2.3E+4 3.9E+3 9.4E+2 4.3E+1 2.8E+0 1.2E+0 4.8E-1 5.7E+1

U-234 7.1E-2

Dose fract 94.4% 93.9% 92.5% 94.5% 94.5% 98.3% 97.1%
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Releases to the environment, which were filtered, consisted mainly of noble fission gases and small 

amounts of iodines. For this analysis, the release was treated as a 1-s release of noble fission gases 

and a small fraction of the halogens produced in the event [10]. Table 4-10 lists the intakes applicable 

at the south end of INTEC and CFA. These intake quantities would be applicable only if the individual 

was in the respective areas on the evening of October 17, 1978.

4.2.2.2.4 HTRE No. 3 Criticality Excursion (IET 13)

IET 13 was characterized as the critical experiments and low-power testing phase of the HTRE No. 3 

reactor configuration for the GE-ANP Program. The low-power and critical experiments began on 

September 8 and ended on November 18, 1958, when a criticality excursion damaged every fuel 

element in the reactor core. As indicated in General Electric Company literature about the test (Linn 

1962), the critical experiments and low-power testing of the assembly created insignificant radiological 

releases in comparison to the release from the 770-MW-s energy excursion that ended this series of 

tests. More information about this incident is included in Peterson (2004).

The released material for this test is consistent with the release modeled in DOE (1991). Table 4-10 

lists the resultant inhaled quantities that would be applicable at the LPTF/Shield Test Pool Facility 

(STPF) location. The intakes are applicable only if the individual was at the LPTF/STPF location on 

November 18, 1958.

4.2.2.3 Releases from Planned Tests

Of the 108 episodic releases analyzed in DOE (1991), only 16 had the potential to affect other INL

facilities [11]. Section 4.2.2.2 describes four of these events – the three criticalities at INTEC and the 

critical excursion in HTRE No. 3. The 12 remaining events are:

1. FEBT A 7. IET 17(B)

2. FEBT B 8. IET 18

3. Fuel Element Cutting Facility (FECF) 9. IET 19(A)

4. IET 14 10. IET 25(A)

5. IET 15(B) 11. IET 25(B)

6. IET 16 12. IET 26(A)

For a given test, if an onsite facility existed between the point at which the test was conducted and the 

affected offsite location, that test was conservatively assumed to have affected an onsite facility or 

facilities. For example, the meteorological conditions that existed during FEBT A at the GRID III 

location dictated that the offsite location of Reno Ranch, adjacent to the northern INL boundary as 

shown on Figure 4-1, be evaluated in DOE (1991). Inspection of the plume passage necessary to 

affect Reno Ranch indicates an impact at the TAN facility. The same conclusion can be drawn for 

FEBT B, but in that case the plume impact on TAN facilities would be less direct because the offsite 

affected location was Birch Creek, which is considerably to the west of Reno Ranch. The FECF filter 

break at INTEC clearly contaminated an area south of INTEC. According to the meteorological 

dispersion at the time of the filter break, the affected offsite location was Frenchman’s Cabin. 

Because CFA is in the straight-line path between INTEC and Frenchman’s Cabin, a radiological 

impact analysis was conducted for CFA. 

Because all other test releases listed above, which originated at the TAN facility, affected two 

locations on the southern boundary [Frenchman’s Cabin or Cerro Grande (shown on Figure 4-1), as 

evaluated in DOE (1991)], they could also have affected INTEC, TRA, SPERT, the south INL Security 

Gate, CFA, or EBR-1. Because TRA and INTEC are essentially the same distance from TAN, 

concentrations at either location were assumed to be the same. The following sections discuss these 

events.
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4.2.2.3.1 Fuel Element Burn Test A

This test supported the GE-ANP Program and provided information on the hypothetical radiological 

release during the crash of a nuclear-powered plane (Brodsky and Beard 1960). This test, with a 

stainless-steel-clad fuel element that had been operated at 20 MW for 120 hours and decayed for 70 

days, began at 2:19 p.m. on March 20, 1957, at GRID III. The fuel element was assembled among 

other airframe parts to simulate an aircraft crash and burned for 2 hours using jet fuel as the burning 

agent. After 2 hours, the fuel element, which was intact, was believed not to have released much 

radioactive material (Brodsky and Beard 1960, p 42). Of the two FEBTs, Test A probably affected

TAN more directly than Test B. Table 4-11 lists best estimates of the radionuclides and the intakes of 

those nuclides at TAN. These intakes would have been received only if the individual was at TAN on 

March 20, 1957. 

4.2.2.3.2 Fuel Element Burn Test B

This test was conducted with a fuel element that had twice the curie content as the first burn test. The 

test began at 6:47 p.m. on the same day as the first burn test (March 20, 1957) and heated the fuel to 

much higher temperatures by supplying oxygen to the fire fueled by thermite, steel wool, and iron 

filings. The fuel element for this test underwent 6.16 × 1021 fissions and a subsequent decay of 250 

days (Brodsky and Beard 1960, p. 37). 

After 90 seconds, most of the fuel element had melted and dispersed. DOE (1991, Figure 3-12) 

shows the trajectory of the plume, and Figure 3-13 in that document shows the corresponding 

dispersion coefficient contours. Interpolation between the contours of 29 and 13,000 × 10-14 hr/m3

provided a concentration for the GE-ANP area in the range of 6 × 10-12 hr/m3

, the value used for 

determining the dose to an individual at TAN. The radionuclides released and the respective curie 

values were the same as those for the DOE (1991) analysis. Table 4-11 lists the resulting intakes for 

TAN. These intakes would have been received only if the individual was at TAN on the evening of 

March 20, 1957. 

4.2.2.3.3 FECF Filter Break

Fuel elements sent to INTEC contained structural components on the ends that were cut off before 

the elements were processed. Cutting these end pieces off and cutting the fuel elements into 

sections before they were sent to Building CPP-601 for processing occurred in the FECF in Building 

CPP-603. During the night of October 29 and early in the morning of October 30, 1958, 

decontamination operations were conducted in the FECF. Acid fumes from the decontamination

operations caused failure of the FECF exhaust filters, which resulted in the release of particulate 

activity to the south of INTEC. 

Approximately 100 Ci of long-half-life particulate radioactive material was released over an area of 

approximately 200 acres (Horan 1959). The released radioactive material and quantities were the 

same as those published in DOE (1991). Table 4-11 lists best-estimate intakes of the radionuclides at

the southern end of INTEC and at the CFA and EBR-1. These intakes would be applicable only if the 

individual was in the area at the time of the release (i.e., during the night of October 29 and the early 

morning of October 30, 1958).

4.2.2.3.4 Initial Engine Test 14

IET 14 was the eighth nuclear test conducted by the GE-ANP Program at TAN. This test was the fifth 

in the HTRE No. 2 reactor configuration. This test series involved the evaluation of the L2A-1 insert 

cartridge. The cartridge contained fueled and unfueled, beryllium-oxygen ceramic tubes. There was 

no coating on the inside surfaces of the fueled tubes (Whitlow et al. 1959). A total of 100.25 hours
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Table4-11. Intakes (Bq/event) for special tests at INL [12].

Date Event

Exposure 

location Sr-89 Sr-90Y-91 Zr-95 Ru-103 Ru-106

I-131 

(elem.) Ce-144 Pr-143

3/20/1957 FEBT A (14:19) TAN 3.9E-2 6.1E-4 4.7E-2 5.0E-2 2.5E-2 1.1E-3 2.0E-2 1.8E-2 1.3E-2

3/20/1957 FEBT B (18:47) TAN 1.3E-2 3.8E-3 2.4E-2 3.2E-2 3.8E-3 4.5E-3 9.0E-9 6.5E-2 2.6E-6

10/29–10/30/58 FECF filter break South INTEC 4.2E-2 4.2E-2 9.1E-2 1.3E-1 8.3E-3 4.3E-2 2.7E-13 5.9E-1 1.8E-7

CFA 1.8E+0 1.8E+0 3.9E+0 5.8E+0 3.6E-1 1.9E+0 1.2E-11 2.6E+1 7.6E-6

EBR-I Area 6.3E-1 6.3E-1 1.4E+0 2.0E+0 1.3E-1 6.6E-1 4.2E-12 9.0E+0 2.6E-6
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was accumulated on the insert fuel cartridge at a maximum insert fuel temperature of approximately 

2,500°

F. The objectives of the test were (1) to evaluate the operational effect of water vapor corrosion 

on fueled beryllium-oxygen tubes operating at a constant reactor mixed mean discharge air 

temperature over approximately 100 hours and (2) to measure the fission product release rate from 

uncoated fueled tubes as a function of temperature and operating time (Whitlow et al. 1959).

Table 4-12 lists the fission products released during IET 14 and the intakes at TRA/INTEC, CFA, and 

EBR-1. An individual would have been exposed to these concentrations only if present at these 

locations between April 24 and May 19, 1959. The above radiological intakes are for the entire

exposure period of 26 days.

4.2.2.3.5 Initial Engine Test 15(B) 

IET 15 was conducted at TAN between May 27 and June 24, 1959. This test involved the evaluation 

of the L2C-1 insert cartridge, which was of the concentric ring design. The fuel sheet was made of a 

chromium-uranium dioxide-titanium core clad with an iron-chromium-yttrium alloy (Evans 1959). Data 

were obtained from this operation to evaluate:

• Endurance capabilities of the advanced metals at a design temperature of 2,000°F for 

extended periods (planned endurance testing to total 120 hours or more)

• The structural and metallurgical integrity of the fuel sheet in this particular cartridge design

• The nature and extent of fuel sheet damage, if any, and the effect on cartridge performance

• The performance potentials of the cartridge

The operation was successfully conducted to accumulate 80.75 hours at an insert extrapolated fuel 

sheet temperature of 2,015o

F. The operation was terminated after 80.75 hours due to a release of 

fission products of such a quantity to indicate fuel sheet rupture of an extent sufficient to warrant 

inspection (Evans 1959).

The insert was visually examined after completion of testing. No damage had occurred to the outer 

fuel sheets of the cartridge, but blisters were observed on the inner fuel sheets. In some instances 

the blisters had ruptured. The fission product release for this test was divided into two periods based 

on a review of effluent monitoring data. The first period – June 3 to 15 – was considered to be an 

operation before the development of significant fuel sheet blisters. June 16 to 24 comprised the 

second period, when effects of blistering were clearly observed (Evans 1959). 

According to the meteorology of the testing period, the second period affected Frenchman’s Cabin. 

Therefore, this analysis addressed the radiological impact on the TRA/INTEC, EBR-1, and CFA. 

Table 4-12 lists the releases of fission products, which correspond with the Part B operation releases

documented in DOE (1991) for the intakes applicable at TRA/INTEC, EBR-1, and CFA. An individual 

would have been exposed to these concentrations and intakes only if present at the locations 

between June 16 and June 24, 1959. The radiological intakes in Table 4-12 are for the entire 9-day

exposure period.

4.2.2.3.6 Initial Engine Test 16 

This was the first power test conducted in HTRE No. 3 to determine the operating characteristics of 

the horizontal core. Because the operation was to determine these characteristics, most operations 

were at low power levels. The total operation occurred between July 28 and October 9, 1959, and

produced only 95 MW-hr of power (Highberg et al. 1959). The fission product release for this test, 
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Table4-12. Intakes (Bq/event) for initial engine tests at INL [13].

Period Test Exposure location Rb-89 Sr-89I-131 (elem.)I-133I-135 Cs-138U-234

4/24–5/19/59 IET 14 TRA/INTEC 7.6E+0 2.3E-2 2.7E+0 1.7E+1 2.4E+1 5.5E+1 4.8E-6

4/24–5/19/59 IET 14 CFA 6.4E+0 2.0E-2 2.3E+0 1.4E+1 2.0E+1 4.6E+1 4.0E-6

4/24–5/19/59 IET 14 EBR-I 5.8E+0 1.7E-2 2.1E+0 1.3E+1 1.8E+1 4.2E+1 3.6E-6

6/16–6/24/59 IET 15(B) TRA/INTEC 4.3E-1 1.1E-3 5.9E-1 2.6E+0 4.2E+0 3.0E+0 5.1E-5

6/16–6/24/59 IET 15(B) CFA 3.6E-1 9.3E-4 5.0E-1 2.2E+0 3.6E+0 2.5E+0 4.3E-5

6/16–6/24/59 IET 15(B) EBR-I 3.3E-1 8.4E-4 4.5E-1 2.0E+0 3.2E+0 2.3E+0 3.9E-5

7/28–10/9/59 IET 16 SPERT 2.7E-2 3.6E-4 6.4E-4 1.3E-2 2.8E-2 9.7E-1 1.2E-7

7/28–10/9/59 IET 16 South gate 2.2E-2 2.8E-4 5.1E-4 1.1E-2 2.2E-2 7.7E-1 9.7E-8

10/12–12/12/59 IET 17(B) TRA/INTEC 2.9E-3 8.2E-3 7.5E-1 3.2E+0 3.3E+0 8.2E-1 7.2E-6

10/12–12/12/59 IET 17(B) CFA 2.4E-3 6.9E-3 6.3E-1 2.7E+0 2.7E+0 6.9E-1 6.1E-6

10/12–12/12/59 IET 17(B) EBR-I 2.2E-3 6.2E-3 5.7E-1 2.4E+0 2.5E+0 6.2E-1 5.5E-6

1/26–2/7/60 IET 18 SPERT 1.2E-2 2.6E-3 8.3E+0 4.7E+1 3.1E+1 4.9E-1 8.1E-6

1/26–2/7/60IET 18 South gate 9.4E-3 2.0E-3 6.6E+0 3.7E+1 2.5E+1 3.9E-1 6.4E-6

2/17–2/29/60 IET 19(A) TRA/INTEC 1.7E-1 8.9E-3 1.3E+0 8.8E+0 1.3E+1 6.4E+0 5.5E-7

2/17–2/29/60 IET 19(A) CFA 1.4E-1 7.5E-3 1.1E+0 7.4E+0 1.1E+1 5.4E+0 4.6E-7

2/17–2/29/60 IET 19(A) EBR-I 1.3E-1 6.8E-3 9.9E-1 6.7E+0 9.9E+0 4.9E+0 4.2E-7

11/22–11/30/60 IET 25(A) TRA/INTEC 9.3E-4 6.1E-4 4.7E-1 5.2E+0 6.7E+0 2.0E-1 2.0E-6

11/22–11/30/60 IET 25(A) CFA 7.8E-4 5.2E-4 3.9E-1 4.3E+0 5.6E+0 1.7E-1 1.7E-6

11/22–11/30/60 IET 25(A) EBR-I 7.1E-4 4.6E-4 3.6E-1 4.0E+0 5.1E+0 1.5E-1 1.5E-6

12/1–12/15/60 IET 25(B) SPERT 1.2E-3 8.0E-4 1.4E+0 8.7E+0 8.3E+0 1.3E-1 3.2E-6

12/1–12/15/60 IET 25(B) South gate 9.5E-4 6.4E-4 1.1E+0 6.9E+0 6.6E+0 1.1E-1 2.5E-6

12/23–12/28/60 IET 26(A) TRA/INTEC 2.2E+0 1.5E-2 2.4E+0 9.5E+0 1.4E+1 2.0E+1 3.3E-5

12/23–12/28/60 IET 26(A) CFA 1.9E+0 1.2E-2 2.0E+0 8.0E+0 1.1E+1 1.7E+1 2.8E-5

12/23–12/28/60 IET 26(A) EBR-I 1.7E+0 1.1E-2 1.8E+0 7.2E+0 1.1E+1 1.5E+1 2.5E-5
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which was modeled the same as that in DOE (1991), was assumed to occur on October 9, 1959. 

Modeling releases for IET 16 involved the preservation of three factors: (1) a burnup of 95 MW-hr, (2) 

a conservative particulate release of 14 Ci (1.5 Ci/hr for 9.5 hours), and (3) a conservative release 

fraction of 3.0 × 10-7 for iodine isotopes, the highest fraction measured during the test. To preserve 

these values and arrive at release fractions for other groups of radionuclides, engineering judgment 

and preliminary data from a few iterations of the RSAC program were used. To meet the stated 

criteria of 1.5 Ci/hr for particulates, the noble gas release fraction was assumed to be 200 times 

greater than that for the iodines, and the release fraction for the solids was assumed to be 10% of the 

halogen release fraction [14]. 

The release for this test series was modeled the same as that for DOE (1991) with Cerro Grande, 

south of the INL boundary, as the location of highest offsite concentration. For this analysis, a 

straight-line trajectory from TAN to Cerro Grande intercepted the SPERT facility and the INL South 

Security Gate just north of the junction of Highways 26 and 20 on the road to CFA, as shown in 

Figure 4-1. Table 4-12 lists the intakes at the respective locations.

4.2.2.3.7 Initial Engine Test 17(B) 

IET 17(B) was performed between October 12 and December 12, 1959. Releases of airborne 

radioactivity occurred between November 2 and December 12 when the reactor operated at power 

levels above 100 kW. The test series involved the evaluation of the L2E-1 insert cartridge (Evans 

1960). Table 4-12 lists the intakes. An individual would have received these intakes only if present at 

the locations between November 2 and December 12, 1959. The radiological intakes in Table 4-12 

are for the entire exposure period of 40 days.

4.2.2.3.8 Initial Engine Test 18

IET 18, conducted between December 23, 1959, and February 8, 1960, and designated as the 

Phase II testing of the HTRE No. 3 engine, was an extension of the test program outlined for IET 16. 

The following is a description of operations during this test series:

The powerplant was transported to Initial Engine Test on December 14, 1959, for final 

checkout in the facility prior to testing. The first engine operation was accomplished 

December 22, 1959, and the first data was taken December 23, 1959, (run No. I-1). 

The reactor was made critical on December 23, 1959, (run No. I-6).

Damage was sustained to the instrumentation circuitry of the powerplant on January 6, 

1960, as a result of failure of the electric aftercooling blowers. The powerplant was 

returned to the Hot Shop on January 7, 1960, for repairs that were completed January 

12, 1960. Testing resumed at Initial Engine Test on January 21, 1960, and finished on 

February 7, 1960.

The initial transfer from chemical operation to full nuclear operation occurred on 

January 26, 1960, (run No. 11-12), and the design conditions for endurance testing 

were initially attained at 11:58 p.m. on January 26, 1960 (run 11-32). A total of 126.42 

hours of operation was achieved at design conditions with a continuous operation of 

64.9 hours at these conditions. Operations were accomplished above 1% of design 

power for a total of 166.5 hours (Highberg et al. 1960).

In relation to effluent monitoring:

Continuous effluent monitoring was maintained to measure and to record the activity 

released to the atmosphere by the powerplant. The maximum output was 8.6 
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curies/hour (measured 10 minutes after release). The total output for the test series 

was 1157 curies (measured 10 minutes after release). The maximum release rate for 

I-131 was approximately 1.5 curies/hour (measured 10 minutes after release). The 

total offsite inhaled and ingested was below measurable amounts during this test 

series. (Highberg et al. 1960)

The release for this test series was modeled the same as DOE (1991) with Cerro Grande, south of the 

INL boundary, as the location of highest concentration. For this analysis, a straight-line trajectory 

from TAN to Cerro Grande intercepted the SPERT facility and the INL South Security Gate. Table

4-12 lists the intakes for the two onsite locations. 

4.2.2.3.9 Initial Engine Test 19(A) 

IET 19(A), conducted between February 9 and April 30, 1960, was a test series in the HTRE No. 2 

reactor to evaluate the L2E-3 insert, which contained fueled and unfueled hexagonal beryllium-oxygen 

ceramic tubes. The tubes were coated on the inside with coextruded zirconia (zirconium dioxide) 

(Pincock 1960). The primary purposes for running the test were:

• To operate the L2E-3 fuel cartridge at peak temperatures of 2,500°F and 2,600°F for 100 

hours or more at each temperature level to evaluate the effectiveness of the zirconium dioxide 

coating against hydrolysis and the release of fission products

• To operate the insert fuel cartridge at various temperature levels at specified intervals during 

the endurance testing to determine fission product release as a function of insert temperature

• To obtain additional information on the effectiveness of an electrostatic precipitator in 

removing fission products from the reactor effluent (Pincock 1960)

Pincock (1960) summarized the estimated total fission product release for the test runs based on spot 

sampling and reported them as 10-minute-decayed curies. The total fission product release reported 

for IET 19(A) was 2,892 Ci. The release for this test was modeled like that for DOE (1991). Table

4-12 lists intakes for TRA/INTEC, EBR-1, and CFA. An individual would have been exposed to these 

intakes only if present at the respective locations between February 17 and February 29, 1960. The 

radiological intakes in Table 4-12 are for the entire exposure period of 13 days.

4.2.2.3.10 Initial Engine Test 25(A) 

IET 25(A), performed between November 15 and December 19, 1960, was an extension of the Phase 

II testing program in IET 18. The test was conducted in the HTRE No. 3 reactor configuration. 

Releases of airborne radioactivity that corresponded to the significant periods of operation were 

assumed to have occurred between November 22 and December 15, 1960. The release at IET 25(A) 

was assumed to have occurred from November 22 through November 30, 1960.

The purposes of test series IET No. 25 were to demonstrate the capabilities of the fuel 

elements above design temperatures and to confirm that the powerplant could achieve 

a full nuclear start as predicted. The reactor went critical on November 14, 1960, and 

the test program was completed on December 19, 1960. (Linn 1962)

Only the following summary of effluent monitoring activities and results was available: 

Continuous effluent monitoring was maintained to measure and record the activity 

released to the atmosphere by the powerplant. The maximum output was 3.4 

curies/hour (measured 10 minutes after release). The total output for the test series 
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was 218 curies (measured 10 minutes after release). The maximum release rate for 

I-131 was approximately 0.7 curies/hour (measured 10 minutes after release). The 

total offsite inhaled and ingested dose was below measurable amounts during this test 

series. (Highberg et al. 1961)

For this analysis, the release was modeled like that in DOE (1991) as a straight-line trajectory such 

that the centerline plume affected TRA/INTEC, EBR-1, and CFA. The intakes for this test are listed in 

Table 4-12. An individual would have received intakes only if present at the locations between 

November 22 and December 15, 1960. The tabulated intakes are for the entire exposure period of 

24 days.

4.2.2.3.11 Initial Engine Test 25(B) 

IET 25, performed between November 15 and December 19, 1960, was an extension of the Phase II 

testing program in IET 18 and the second part of IET 25 (Linn 1962; Highberg et al. 1961). The 

releases for the test correspond to the significant periods of operation with IET 25(B) releases 

occurring from December 1 to December 15, 1960. Of the total release for IET 25, 76% was 

assumed to have been released during the IET 25(B) operation.

The releases for this test were the same as those modeled for DOE (1991) with a straight-line 

trajectory from TAN to SPERT and the INL South Security Gate. Table 4-12 lists the intakes that 

correspond to the two onsite locations. An individual would have received intakes only if present at 

the locations between December 1 and December 15, 1960. The tabulated radiological intakes are 

for the entire exposure period of 16 days.

4.2.2.3.12 Initial Engine Test 26(A) 

IET 26(A), conducted in HTRE No. 2, was performed between December 22, 1960, and March 31, 

1961 (Foster et al. 1961). Releases of airborne activity for the total test were assumed to have 

occurred between December 23, 1960, and March 30, 1961, when the reactor operated at power 

levels above 120 kW. Releases for the IET 26(A) operation occurred from December 23 to 28, 1960. 

The insert being tested was the L2E-6 cartridge, which consisted of fueled and unfueled ceramic 

beryllium-oxide hexagonal tubes coated on the inner surface with zirconium dioxide (Foster et al.

1961).

The airborne release model was consistent with the model of DOE (1991) with an assumed straightline trajectory between TAN, TRA/INTEC, EBR-1, and CFA. The intakes are listed in Table 4-12 for 

TRA/INTEC, EBR-1, and CFA. An individual would have been exposed to these intakes only if 

present at the locations between December 23 and December 28, 1960. The tabulated intakes are 

for the entire exposure period of 6 days.

4.3 EXTERNAL DOSE

External radiation dose at a facility can be created by direct radiation from two sources: (1) direct 

beta/gamma radiation from the facility or (2) gaseous effluents released from the facility or from 

adjacent facilities. In general, direct beta/gamma radiation from the facility increases with time 

because the general contamination of the area increases. In addition, as a facility ages, radioactive 

sources tend to accumulate at the facility, which causes the general background to increase with time. 

A responsible H&S organization observes and curbs such trends to prevent personnel exposures from 

increasing unnecessarily [15]. An excellent example of another aspect that can cause facility 

background to increase is the operational tests at the IET area at TAN. During an IET, such as one of 

those described in Section 4.2.2.3 where fuel damage created a significant release to the 

environment, the 76-in. duct from the HTRE engine to the stack was a significant radiation source. 
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 42 of 67

That is why operational facilities for that area were heavily shielded and personnel were required to be 

inside a shielded facility during a test. The following sections discuss facility fenceline film badge and 

TLD data that recorded doses from gaseous fission product releases that had the potential for 

personnel exposure. More information on these two subjects is provided in Peterson (2004).

4.3.1

Before 1970, many film badge or TLD measurements were made inside the INL boundary. During the 

IET period at TAN (1956 to 1961), many film badges were placed along the highways that triangulated 

the IET area and along highways at the south end of the site. The badges were initially retrieved and 

read once a month. The frequency increased to 6 weeks in 1962 and changed back to monthly in 

1963. Film badges were used through the first 9 months of 1966, and TLDs were used from that time

forward (AEC 1967a, p 8). Beginning in 1967, TLDs were changed on a semiannual basis. 

Significant readings during the film badge period showed that the maximum badge reading increased 

by only a factor of 2 or 3 above background. However, the location of the badge with the increased 

reading was not identified. More information and film badge data for this early period are provided in 

Peterson (2004). The detection limit for the film badge reading was often quoted as 10 mrem for both 

a beta reading and a gamma reading (Dodd 1963) and that for the TLD was quoted as being 10 mrem 

when it was first used. With the annual background radiation field at INL before the start of operations 

measured at 100 to 150 mrem/yr, the monthly value of 8 to 13 mrem is at the detection limit of the film 

badge or the TLD [16]. Therefore, the uncertainty for monthly change-outs is higher than that for less 

frequent change-outs. 

Facility Fenceline Annual Doses

Between the latter part of 1970 and the latter part of 1972, facility fenceline monitoring, facility 

locations, and facility arrangements had been established to provide a consistent location for TLD 

monitors (Walker 1973a, p. 1-13). Although there have been minor facility changes since 1972, an 

effort was made to choose TLD locations that have been consistent from 1972 to the present. In most 

cases, a facility change did not result in location changes for TLDs. The TRA facility did have minor 

changes over time but the majority of the TLD locations were consistent. TRA initially had an “L” 

shape, but in 1992, the facility plot changed to a square. No film badge or TLD information is 

available for the EBR-I and BORAX locations when they were operational. 

There are two areas at TRA where TLDs were used especially to track gamma fields for work 

planning: (1) the North Storage Area (NSA), and (2) the TRA Warm Waste Ponds (WWPs). The NSA 

was designated for storage of mildly activated or contaminated reactor hardware. It was just outside 

and north of the TRA perimeter, but still within the area fence and away from normally inhabited 

buildings or areas. TLDs 7, 8, and 9 were placed around the NSA to track area radiation fields due to 

the irradiated material. The TRA WWPs were outside the TRA to the east. The elevation of the 

ponds was at least 5 ft below the elevation of the TRA facility. TLDs 3, 4, and 5 were placed around 

the WWPs to track radiation fields from the ponds for personnel control and exposure reasons. In the 

original attempt to characterize the TRA radiation field, TLDs 3, 4, 5, 7, 8, and 9 were excluded. The 

final characterization used TLDs 1, 7, 12, and 13, which increased the gamma field for TRA by a 

factor of 3 to 4 above the initial value. Row 2 of Table 4-13 lists area TLD numbers.

From 1972 through 1983, facility fenceline TLD measurements, made on a 6-month basis with five

TLDs at each facility position, are available in the quarterly Environmental Monitoring Data reports 

(Walker 1972, 1973a,b, 1974a,b, 1975a,b,c; Dahl 1975; Sill 1976; Bills 1976; Williamson 1976a,b, 

1977a,b,c, 1978a,b; DOE 1978b,c,d, 1979b,c,d,e, 1980b,c,d,e, 1981b,c,d,e, 1982b,c,d,e, 1983b,c,d,e, 

1984). Figure 4-7 (from Williamson 1976a) shows that uncertainty can vary from less than 10% to 

20% for a given set of readings. At each of the 34 monitoring locations, there are normally five TLDs 

for a potential 170 readings for a 6-month period. For this particular 2-year set of data, 1.5% of the 

136 readings were assigned a 2-sigma uncertainty of 16 to 20% and 18.4% of the readings were 
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2 Effective Date: 01/08/2010 Page 43 of 67

Table 4-13. INL facility fence direct gamma values (TLD–background) (mR) [17].

Year

ARA 

I & II SPERT

TANTSFTANLOFTTANLPTF CFA TRA INTEC RWMC EBR-II TREAT Background

TLDs used

1, 2, 3 All All All All All 1, 7, 12, 13 2, 11, 12, 13a 13, 15, 17, 19 All All

1952–72 226 12 42 10 12 52 438 446 32 36 18 100–150

1973 86 21 41 17 12 53 306 405 32 37 19 121

1974 162 48 8 7 0 59 320 627 370 35 17 123

1975 114 16 29 11 7 17 195 357 265 32 8 118

1976 66 27 22 15 12 20 140 311 155 56 50 113

1977 41 5 0 4 4 1 137 318 189 22 0 132

1978 52 12 2 9 4 7 143 251 106 56 2 129

1979 63 18 10 17 7 14 159 236 65 59 5 113

1980 65 17 8 19 13 18 251 203 57 51 12 119

1981 63 18 8 17 10 14 231 255 42 28 9 118

1982 50 26 6 12 6 10 163 124 42 20 12 117

1983 78 17 23 26 19 18 174 141 50 24 10 115

1984 80 19 11 19 12 15 205 181 48 31 13 124

1985 80 19 11 19 12 15 205 181 48 31 13 124

1986 80 19 11 19 12 15 205 181 48 31 13 124

1987 80 19 11 19 12 15 205 181 48 31 13 124

1988 80 19 11 19 12 15 205 181 48 31 13 124

1989 80 19 11 19 12 15 205 181 48 31 13 124

1990 80 10 10 11 9 11 28 39 27 19 13 124

1991 80 10 10 11 9 11 28 39 27 19 13 124

1992 80 10 10 11 9 11 28 39 27 19 13 124

1993 77 19 18 23 15 15 48 37 24 28 16 111

1994 69 0 0 0 0 0 24 28 25 15 3 130

1995 91 6 4 8 2 11 31 43 42 17 7 116

1996 52 4 14 0 0 13 28 49 40 22 21 129

1997 46 10 3 8 0 9 29 44 17 16 16 128

1998 62 8 0 5 0 12 25 31 20 0 11 131

1999 49 13 0 0 0 5 10 38 22 13 13 122

2000 28 16 7 16 8 19 40 55 61 25 26 129

2001 31 3 0 0 0 0 27 32 25 0 3 140

2002 41 11 0 0 0 9 34 54 33 18 39 120

2003 43 21 0 8 0 11 31 15 14 20 11 117

2004 24 11 6 7 1 7 26 13 19 23 10 118

2005 22 13 3 8 4 12 23 20 10 26 24 120

2006 4 7 0 0 0 7 21 36 3 6 2 122

2007 2 8 0 0 0 1 23 40 6 6 2 130

a. All after 2005.
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assigned a 2-sigma uncertainty of 11% to 15%. However, 80% of the 136 values ascribed for the 34 

locations over the 2-year period have a 2-sigma uncertainty of less than 10% [18].

Figure 4-7. Example of onsite TLD monitoring data (Williamson 1976a, p. 38). 

To supply facility values for the period from 1952 to 1972, the highest 6-month value from April 1972 

to April 1973 for a facility was multiplied by 2 and applied to each year between 1952 and 1972 [19]. 

(Not all the listed facilities had begun operation or even existed in 1952). Facility fenceline TLD 

measurements could not be found for 1984 through 1992, but for 1993 and beyond such 

measurements were included in the EMRs (Mitchell 1994; Mitchell, Peterson, and Hoff 1995; Mitchell 

et al. 1996, 1997; Evans et al. 1998; Saffel et al. 2000; Stoller 2002a,b,c, 2003). For the period from 
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 45 of 67

1984 to 1992 when TLD measurements were missing, reasonable interpolations were used to provide 

the missing values. In addition, background TLD measurements that correspond with the facility 

fenceline TLD measuring periods were recorded in the EMRs. All reduced facility fenceline TLD data 

(facility fenceline data minus background) in the EMRs are listed in Table 4-13 (AEC 1971, 1972, 

1973,1974; ERDA 1975, 1976, 1977b; DOE 1978a, 1979a, 1980a, 1981a, 1982a, 1983a; Hoff, Chew, 

and Dickson 1984; Mitchell 1994; Mitchell, Peterson, and Hoff 1995; Mitchell et al. 1996, 1997; Evans 

et al. 1998; Saffel et al. 2000; Stoller 2002a,b,c, 2003, 2007, 2008). A more detailed discussion of the 

data is included in Peterson (2004).

4.3.2

INL facility air immersion (beta-gamma) doses could be calculated from the noble gas and halogen 

portions of the operational releases and, if applicable, from the noble gas portion of the applicable 

episodic releases. This calculation should be unnecessary because these releases would be 

recorded in the fenceline TLD doses listed in Table 4-13.

Facility Air Immersion Doses

4.4 UNCERTAINTY

A detailed discussion of the derivation of airborne releases for operational conditions and episodic 

events is provided in DOE (1991).

Operational Releases. Discussions with the authors of DOE (1991) suggest that operational 

releases, which were monitored, could be low by a factor of not more than 2 [20]. When annual 

normalized ground-level concentration values are applied to operational releases, the uncertainty 

could be increased. However, in considering this increased uncertainty, it is interesting to note the 

facility air-monitoring results in the annual EMRs. In each case, the facility air concentration is 

compared to that for a distant community, usually Idaho Falls. The concentration is normally 

indistinguishable from the concentration for the distant community, as discussed in Section 4.2.1 [21]. 

Episodic Releases. As described in DOE (1991), episodic releases are maximum reasonable values

based on the amount of material available to be released and the conditions of the test. For such 

releases, the inhaled quantities (in becquerels) were maximized by assuming the downwind exposed 

individual was subjected to the plume centerline concentration for the total time, night and day in most 

cases, of the release. In spite of the effort to be reasonably conservative in exposure estimates, 

some of the authors of DOE (1991) have stated that the release considered for a particular episodic 

event might be low by as much as a factor of 3 [22].

A concerted effort has been made to reduce the number of radionuclides involved in releases for 

episodic events. Overall, the mix of radionuclides for all the episodic events is complicated by the 

type (aged versus fresh) and the relative quantities of each. When viewed together, the episodic 

events can be categorized into three categories: (1) criticalities that involve fresh fission products that 

have relatively short half-lives in comparison to radionuclides released from the FEBTs, for example;

(2) releases involving long-half-life aged fission products (FEBTs and the release from the FECF Filter 

Break); and (3) releases from the remaining IETs that released short-half-life radionuclides, which are 

generally characterized as fresh fission products, and long-half-life radionuclides, which are 

characterized as aged fission products. The latter category is unique to the GE-ANP Program 

because of the direct-to-air conversion nature of the tests. Therefore, within these categories, the 

number of radionuclides has been reduced to the number that preserves 95% of the original dose 

calculated for that location [23]. 

In 1988, a minor contamination incident occurred at the RWMC that led to minor inhalation exposures 

to a few individuals, but that incident has not been incorporated into the environmental doses. The 
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following paragraph from the 1988 EMR (Hoff, Mitchell, and Moore 1989) summarizes site monitoring

information for this incident:

In early spring of 1988, waste boxes which had been stored on an outdoor asphalt pad 

at the RWMC (some since 1986) were to be moved indoors. A box which had been 

stored outside was discovered to be breached with some contamination spread. 

Further investigation of approximately 500 boxes on the pad found three other boxes 

that had been breached. Since these boxes had been stored on the pad for more than 

a year and were known to contain primarily Pu-238, it is now assumed that 

concentrations of that radionuclide reported on RWMC air filters in 1986 and 1988 

were due to the small releases from those boxes. The highest RWMC concentration of 

Pu-238 in 1986 was 0.06% of the derived concentration guide and in 1988 was 0.04% 

of the guide. Cleanup and contamination controls were instituted at RWMC, and the 

entire stack of boxes was covered in November 1988 as an interim measure to control 

the spread of contamination prior to a permanent solution to the problem. 

Film Badge and TLD Measurements. As discussed in Section 4.3.1, the uncertainty of individual 

measurements made with film badges and TLDs can be as high as ±100% depending on the 

frequency of change-out (i.e., once per month, which was generally the case with film badges). The 

data for 1952 to 1972 in Table 4-13 are based on the highest 6-month TLD values from 1972 for the 

facility. Although the GE-ANP IETs were conducted in the late 1950s and early 1960s (IET 26, the 

last, ended on March 31, 1961), tests with planned releases were administratively and 

meteorologically controlled so the airborne effluent traveled to the northeast over the monitoring grid 

and so that adjacent facilities were not affected. In spite of these controls, the 1952 to 1962 values for 

the TAN areas (TSF, LOFT, and LPTF) could be low by a factor of 3 [24]. However, after 1967 when 

facility fenceline measurements were routinely made with five TLDs at a given location, the 

uncertainty is generally ascribed at less than 10% or about 13 mrem. Less than about 20% of the 

time, these measurements have an ascribed level of uncertainty as high as 20% [25].

Dose reconstruction for an individual whose location is unknown should use the intakes listed in

Table 4-3 (CFA) and the exposures for the area with the highest annual dose for each year of 

employment listed in Table 4-13. The suggested values maximize the resultant individual dose. 

4.5 ATTRIBUTIONS AND ANNOTATIONS 

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations to assist in the process of worker dose reconstruction. These 

callouts are listed in this section with information that identifies the source and justification for each 

item. Conventional references are provided in the next section that link data, quotations, and other 

information to documents available for review on the Oak Ridge Associated Universities (ORAU) 

Team servers.

Norman Rohrig served as the initial Document Owner for this document. Mr. Rohrig was previously 

employed at INL and his work involved management, direction, or implementation of radiation 

protection and health physics program policies, procedures, or practices related to atomic weapons 

activities at the site. This revision has been overseen by a new Document Owner, who is fully 

responsible for the content of this document, including all findings and conclusions. In all cases 

where such information or prior studies or writings are included or relied on by the Document Owner, 

those materials are fully attributed to the source.

Henry Peterson served as a Subject Expert for this document. Mr. Peterson was previously employed 

at INL and his work involved management, direction, or implementation of radiation protection and 

health physics program policies, procedures, or practices related to atomic weapons activities at the 
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 47 of 67

site. As mentioned above, this revision has been overseen by a new Document Owner who is fully 

responsible for the content, including all findings and conclusions. In all cases where such 

information or prior studies or writings are included or relied upon by Mr. Peterson, those materials 

are fully attributed to the source.

[1] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

Although major events at one facility have sometimes been measured at other nearby 

facilities, they have never necessitated emergency measures.

[2] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

Tests were conducted over the GRID III when the meteorological conditions guaranteed the 

effluent would not affect other INL facilities. The monitoring grid was directed to the northeast, 

a direction that was not toward any other facility at the INL Site.

[3] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

Figure 4-3 shows a concentration isopleth of about 5 for Idaho Falls compared with an isopleth 

over 100 for all site facilities.

[4] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

This is a recommendation to use the absorption type that is predominant based on information 

on chemical form and solubility type found in Peterson (2004). 

[5] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

This is a recommendation to use the absorption type that will give the largest dose. 

[6] Peterson, Henry. ORAU Team. Site Expert. September-December 2003. 

This limited list of radionuclides provides more than 95% of the effective dose equivalent 

based on International Commission on Radiological Protection (ICRP) Publication 68 (ICRP 

1994). The complete list of radionuclides from Peterson (2004) was multiplied by the effective 

dose coefficient for inhalation of 5-μm activity median aerodynamic diameter (AMAD) from 

ICRP Publication 68 and then added together. In steps, nuclides were eliminated based on 

effective dose being less than 0.1% of the total, then 0.3%, and then 0.4%. The fraction of 

effective dose equivalent lost is less than 5%, but the activity lost is considerably more.

[7] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

An example of such analysis is AEC (1956).

[8] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

The default GSD of 3 has been confirmed by telephone conversation and e-mail from 

C. Bloom of the ORAU Team to N. Rohrig on November 9, 2006.

[9] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

This limited list of radionuclides provides more than 95% of the effective dose equivalent 

based on ICRP Publication 68 (ICRP 1994). The complete list of radionuclides from Peterson 

(2004) was multiplied by the effective dose coefficient for inhalation of 5-μm AMAD from ICRP 

Publication 68 and then added together. In steps, nuclides were eliminated based on effective 

dose being less than 0.1% of the total, then 0.3%, and then 0.4%. The fraction of effective 

dose equivalent lost is less than 5%, but the activity lost is considerably more.

[10] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

These values are generally consistent with reactor safety licensing assumptions.
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 48 of 67

[11] Peterson, Henry. ORAU Team. Site Expert. September-December 2003. 

When the analysis of DOE (1991) was completed, the meteorological diffusion trajectories 

were reviewed to determine which INL facilities were affected. The result of these trajectory 

reviews was that only 16 releases had the potential to affect other INL facilities.

[12] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

This limited list of radionuclides provides more than 95% of the effective dose equivalent 

based on ICRP Publication 68 (ICRP 1994). The complete list of radionuclides from Peterson

(2004) was multiplied by the effective dose coefficient for inhalation of 5-μm AMAD from ICRP 

Publication 68 and then added together. In steps, nuclides were eliminated based on effective 

dose being less than 0.1% of the total, then 0.3%, and then 0.4%. The fraction of effective 

dose equivalent lost is less than 5%, but the activity lost is considerably more.

[13] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

This limited list of radionuclides provides more than 95% of the effective dose equivalent 

based on ICRP Publication 68 (ICRP 1994). The complete list of radionuclides from Peterson 

(2004) was multiplied by the effective dose coefficient for inhalation of 5-μm AMAD from ICRP 

Publication 68 and then added together. In steps, nuclides were eliminated based on effective 

dose being less than 0.1% of the total, then 0.3%, and then 0.4%. The fraction of effective 

dose equivalent lost is less than 5%, but the activity lost is considerably more.

[14] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

These values are generally consistent with reactor safety licensing assumptions.

[15] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

As a facility ages, dirt and waste, including radioactive material, generally accumulate in the 

facility. A good example of such accumulation is at a foundry or machine shop. 

Knowledgeable management will remove these materials and wastes as they accumulate; if 

not, the general background radiation, as well as clutter of the facility, will increase with time.

[16] Peterson, Henry. ORAU Team. Site Expert. September-December 2003. 

The uncertainty of 10 mrem is roughly the same as the expected result of 8 to 13 mrem. 

[17] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

See the references listed in the last paragraph of Section 4.3.1.

[18] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

This comes from counting the values in Figure 4-7.

[19] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

This explains how missing data were generated.

[20] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

Discussion between Henry Peterson, Doug Wenzel (INL), and Richard Dickson (INL) in June 

2003.

[21] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

These results are shown in the EMRs, for example, Walker (1972).

[22] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

Discussion between Henry Peterson, Doug Wenzel (INL), and Richard Dickson (INL) in June 

2003.
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 49 of 67

[23] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

This limited list of radionuclides provides more than 95% of the effective dose equivalent 

based on ICRP Publication 66 (ICRP 1994). The complete list of radionuclides from Peterson 

(2004) was multiplied by the effective dose coefficient for inhalation of 5-μm AMAD from 

Appendix B of ICRP Publication 66 and then added together. In steps, nuclides were 

eliminated based on effective dose being less than 0.1% of the total, then 0.3%, and then 

0.4%. The fraction of effective dose equivalent lost is less than 5%, but the activity lost is 

considerably more.

[24] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

This follows from the discussion of episodic releases in this uncertainty section. Such 

releases primarily affected these areas.

[25] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

This is based on Figure 4-7, which is a typical set of results.
Document No. ORAUT-TKBS-0007-4 Revision No. 02 Effective Date: 01/08/2010 Page 50 of 67

REFERENCES

AEC (U.S. Atomic Energy Commission), 1956, Analysis of Stack and Field Samples During IET 

Operations, Idaho Operations Office, National Reactor Testing Station, Idaho. [SRDB Ref ID: 

8393]

AEC (U.S. Atomic Energy Commission), 1960, Environmental Monitoring Data for the National 

Reactor Testing Station, Calendar Year 1959 and 1st Quarter of 1960, Idaho Operations 

Office, National Reactor Testing Station, Idaho. [SRDB Ref ID: 1286]

AEC (U.S. Atomic Energy Commission), 1961a, Environmental Monitoring Data for the National 

Reactor Testing Station, October through December 1960, Idaho Operations Office, National 

Reactor Testing Station, Idaho. [SRDB Ref ID: 1290]

AEC (U.S. Atomic Energy Commission), 1961b, Environmental Monitoring Data Annual Summary for 

National Reactor Testing Station 1960, Idaho Operations Office, National Reactor Testing 

Station, Idaho. April 24. [SRDB Ref ID: 1290, p. 4]

AEC (U.S. Atomic Energy Commission), 1962a, Environmental Monitoring Data (Report No. 9), 4th 

Quarter of 1961 and Annual Summary for the National Reactor Testing Station, Idaho 

Operations Office, National Reactor Testing Station, Idaho. [SRDB Ref ID: 1294]

AEC (U.S. Atomic Energy Commission), 1962b, Environmental Monitoring Data (Report No. 10) for 

the National Reactor Testing Station, First and Second Quarter 1962, Idaho Operations Office, 

National Reactor Testing Station, Idaho. [SRDB Ref ID: 1298]

AEC (U.S. Atomic Energy Commission), 1963, Environmental Monitoring Data (Report No. 12) for the 

First and Second Quarters, 1963, Idaho Operations Office, National Reactor Testing Station, 

Idaho. [SRDB Ref ID: 1301]

AEC (U.S. Atomic Energy Commission), 1964a Environmental Monitoring Data (Report No. 13) for the 

Third and Fourth Quarter and Annual Summary - 1963, Idaho Operations Office, National 

Reactor Testing Station, Idaho. [SRDB Ref ID: 1301, p. 13]

AEC (U.S. Atomic Energy Commission), 1964b, Environmental Monitoring Report (Report No. 14) for 

the National Reactor Testing Station, First and Second Quarter 1964, Idaho Operations Office, 

National Reactor Testing Station, Idaho. [SRDB Ref ID: 1313]

AEC (U.S. Atomic Energy Commission), 1965a Environmental Monitoring Report No. 15 for the 

National Reactor Testing Station, Third and Fourth Quarter 1964 and Annual Summary, Idaho 

Operations Office, National Reactor Testing Station, Idaho. [SRDB Ref ID: 1313, p. 7]

AEC (U.S. Atomic Energy Commission), 1965b, Environmental Monitoring Report No. 16, First and 

Second Quarter 1965, Idaho Operations Office, National Reactor Testing Station, Idaho. 

[SRDB Ref ID: 1314]

AEC (U.S. Atomic Energy Commission), 1966a, Environmental Monitoring Report No. 17*, Third and 

Fourth Quarter 1965 and Annual Summary, Idaho Operations Office, National Reactor Testing 

Station, Idaho. [SRDB Ref ID: 1314, p. 5]

AEC (U.S. Atomic Energy Commission), 1966b, Environmental Monitoring Report No. 18*, First & 

Second Quarter 1966, Idaho Operations Office, National Reactor Testing Station, Idaho. 

[SRDB Ref ID: 1324]
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AEC (U.S. Atomic Energy Commission), 1967a, Environmental Monitoring Report No. 19*, Third and 

Fourth Quarter and Annual Summary 1966, Idaho Operations Office, National Reactor Testing 

Station, Idaho. [SRDB Ref ID: 1324, p. 5]

AEC (U.S. Atomic Energy Commission), 1967b, Environmental Monitoring Report No. 20*, First and 

Second Quarter 1967, Idaho Operations Office, National Reactor Testing Station, Idaho. 

[SRDB Ref ID: 1344]

AEC (U.S. Atomic Energy Commission), 1968, Environmental Monitoring Report No. 21*, July -

December and Annual Summary 1967, Idaho Operations Office, National Reactor Testing 

Station, Idaho. [SRDB Ref ID: 1344, p. 5]

AEC (U.S. Atomic Energy Commission), 1968b, Environmental Monitoring Report No. 22 JanuaryJune 1968, Idaho Operations Office, National Reactor Testing Station, Idaho. [SRDB Ref ID: 

1353]

AEC (U.S. Atomic Energy Commission), 1969, Environmental Monitoring Report No. 23*, July -

December and Annual Summary - 1968, Idaho Operations Office, National Reactor Testing 

Station, Idaho. [SRDB Ref ID: 1353, p. 7]

AEC (U.S. Atomic Energy Commission), 1969b, Environmental Monitoring Report No. 24, January -

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Williamson, M. M., 1978a, “INEL Site Environmental Monitoring Data - Third Quarter 1977,”

attachment to untitled letter to R. W. Kiehn (EG&G Idaho), U.S. Department of Energy, Idaho 

Operations Office, Idaho National Engineering Laboratory, Idaho Falls, Idaho, January 11.

[SRDB Ref ID: 1527, p. 82]

Williamson, M. M., 1978b, "INEL Site Environmental Monitoring Report - Fourth Quarter 1977," 

attachment to untitled letter to R. W. Kiehn (EG&G Idaho), Idaho Center for Radiological and 

Environmental Sciences, March 28. [SRDB Ref ID: 1530]

Yanskey, G. R., E. H. Markee, Jr., and A. P. Richter, 1966, Climatography of the National Reactor 

Testing Station, IDO-12048, U.S. Department of Commerce, Environmental Science Services 

Administration, Air Resources Field Research Office, Idaho Falls, Idaho, January. [SRDB Ref 

ID: 12736]
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GLOSSARY

absorption 

In external dosimetry, process in which radiation energy is imparted to material. In internal 

dosimetry, movement of material to blood regardless of mechanism. 

absorption type 

Categories for materials according to their rate of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively in the 

respiratory tract (slow solubilization). Also called solubility type.

activity

Amount of radioactivity. The International System unit of activity is the becquerel 

(1 disintegration per second); the traditional unit is the curie [37 billion (3.7 × 1010) becquerels].

activity median aerodynamic diameter (AMAD) 

Diameter of a unit density sphere with the same terminal settling velocity in air as that of the 

aerosol particle whose activity is the median for the entire aerosol. 

alpha radiation 

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic

charge of +2. 

becquerel (Bq) 

International System unit of radioactivity equal to 1 disintegration per second; 1 curie equals

37 billion (3.7 × 1010) Bq. 

beta dose 

Designation (i.e., beta) on some records for external dose from beta and less-energetic X-ray

and gamma radiation, often for shallow dose or dose to the lens of the eye. 

beta radiation 

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

breeder reactor 

Nuclear reactor that makes more new fissionable material than it consumes. 

contamination 

Radioactive material in an undesired location including air, soil, buildings, animals, and 

persons.

decontamination 

Reduction or removal of radioactive material from a structure, area, object, or person. 

Decontamination can occur through (1) treating the surface to remove or decrease the 

contamination or (2) allowing natural radioactive decay to occur over a period of time.
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dose 

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of roentgens, rads, reps, or grays.

dose equivalent

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose.

dose reconstruction (DR) 

Process of analyzing the available information including evaluation of historical methods and 

data to estimate the dose a person could have received from one or more radiation exposures.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See film dosimeter and thermoluminescent dosimeter.

dosimetry 

Measurement and calculation of internal and external radiation doses.

effective dose equivalent (EDE) 

Average of the dose equivalents weighted for the susceptibility of harm to different tissues or 

organs in units of rem or sievert. See dose and weighting factor. 

effluent 

Liquid or gaseous waste released to the environment.

Energy Employees Occupational Illness Compensation Program Act of 2000, as 

amended (EEOICPA; 42 U.S.C. § 7384 et seq.) 

Law that provides for evaluation of cause and potential compensation for energy employees 

who have certain types of cancer.

exposure 

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

external dose 

Dose received from radiation emitted by sources outside the body.

external radiation 

Radiation from sources outside the body.

film

(1) In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. See film dosimeter. (2) X-ray film.

film badge 

See film dosimeter.
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film dosimeter 

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

fissile 

Capable of undergoing nuclear fission by capturing thermal (slow) neutrons. The three 

primary fissile materials are 233U, 235U, and 239Pu. Fissile generally indicates spontaneous 

fission.

fission 

Splitting of the nucleus of an atom (usually of a heavy element) into at least two other nuclei 

and the release of a relatively large amount of energy. This transformation usually releases 

two or three neutrons. 

fuel 

Fissile material in a nuclear reactor that is the source of the chain reaction of neutrons that 

produce fission and release energy as heat and other electromagnetic radiation.

gamma radiation 

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are 

identical to X-ray photons of high energy; the difference is that X-rays do not originate in the 

nucleus. 

gamma ray, particle, or photon (γ) 

See gamma radiation.

geometric mean

The nth root of the product of all the members of a set of positive numbers, where n is the 

number of members. 

geometric standard deviation (GSD) 

In probability theory and statistics, the spread of a set of numbers whose preferred average is 

the geometric mean.

half-life

Time in which half of a given quantity of a particular radionuclide disintegrates (decays) into 

another nuclear form. During one half-life, the number of atoms of a particular radionuclide 

decreases by one half. Each radionuclide has a unique half-life ranging from millionths of a 

second to billions of years.

intake 

Radioactive material taken into the body by inhalation, absorption through the skin, injection, 

ingestion, or through wounds. 

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, high-
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speed electrons, high-speed protons, photoelectrons, Compton electrons, positron/negatron 

pairs from photon radiation, and scattered nuclei from fast neutrons. See alpha radiation, beta 

radiation, gamma radiation, neutron radiation, photon radiation, and X-ray radiation.

isotope 

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties. 

monitoring 

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 

monitoring includes internal and external dosimetry including interpretation of the 

measurements. 

National Council on Radiation Protection and Measurements (NCRP) 

Private U.S. public service organization chartered by the U.S. Congress to formulate and 

disseminate information, guidance, and recommendations on radiation protection and 

measurements.

National Institute for Occupational Safety and Health (NIOSH) 

U.S. agency responsible for dose reconstruction under the Energy Employees Occupational 

Illness Compensation Program. Part of the Centers for Disease Control and Prevention, which 

is part of the U.S. Department of Health and Human Services, NIOSH is the Federal agency 

responsible for conducting research and making recommendations for the prevention of 

work-related injury and illness.

neutron (n) 

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen. See element.

neutron radiation 

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons. See neutron.

noble gas 

Gaseous chemical element that does not easily react chemically with other elements (i.e., they 

are inert). 

nuclide 

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

photon 

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz. 
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photon radiation 

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays. 

probability of causation (POC)

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Act, the percent likelihood, at the 99th percentile, that a worker incurred a 

particular cancer from occupational exposure to radiation.

radiation

As used in this document, ionizing radiation.

radiation monitoring

See monitoring.

radiation field 

Area around a source into which radiation dissipates. 

radiation source 

(1) Any object or substance that emits radiation. (2) Package of radioactive material 

constructed to have specific radiation properties used for example for medical purposes or to 

calibrate dosimeters.

radiation type 

Of importance in health physics, alpha, beta, photon (X- or gamma), or neutron radiation.

radioactive 

Of, caused by, or exhibiting radioactivity.

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei. 

radionuclide 

Radioactive nuclide. See radioactive and nuclide. 

reactor 

Device in which a fission chain reaction occurs under controlled conditions to produce heat or 

useful radiation for experimental purposes or to generate electrical power or nuclear fuel. 

reactor vessel 

Structure enclosing the fuel elements, control elements, coolant piping, and other structures 

that support the reactor core. 

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.
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roentgen (R) 

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0°C and standard atmospheric pressure. An 

exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for 

higher energy photons (generally greater than 100 kiloelectron-volts). 

routine monitoring 

Monitoring carried out at regular intervals during normal operations.

shallow dose equivalent 

Dose equivalent in units of rem or sievert at a depth of 0.07 millimeters (7 milligrams per 

square centimeter) in tissue equal to the sum of the penetrating and nonpenetrating doses.

sievert

International System unit for dose equivalent, which indicates the biological damage caused 

by radiation. The unit is the radiation value in gray (equal to 1 joule per kilogram) multiplied by 

a weighting factor for the type of radiation and a weighting factor for the tissue; 1 Sv equals 

100 rem.

solubility type 

See absorption type.

technical basis document (TBD) 

Oak Ridge Associated Universities Team document that describes a feature of a site in 

relation to the Energy Employees Occupational Illness Compensation Program. See site 

profile.

thermoluminescence 

Property that causes a material to emit light as a result of heat.

thermoluminescent dosimeter (TLD) 

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated by radiation, release the stored energy as light. The measurement of this 

light provides a measurement of absorbed dose. 

thermoluminescent dosimeter chip 

Small block or crystal of lithium fluoride in a thermoluminescent dosimeter. A TLD-600 

dosimeter contains a chip made from more than 95% 6

Li for neutron radiation detection, and a 

TLD-700 dosimeter contains a chip made from more than 99.9% 7

Li for photon and beta 

radiation detection. Also called crystals.

U.S. Atomic Energy Commission (AEC) 

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The U.S. Energy Research and Development 

Administration and the U.S. Nuclear Regulatory Commission assumed separate duties from 

the AEC in 1974. The U.S. Department of Energy succeeded the U.S. Energy Research and 

Development Administration in 1979.
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U.S. Department of Energy (DOE) 

Federal agency created in 1979 that assumed, from the Energy and Research Development 

Administration, the responsibility for development of new reactors along with production and 

use of nuclear materials and weapons by the Federal government.

U.S. Department of Labor (DOL) 

Agency that oversees compliance with Federal labor laws as well as collects labor-related 

information. DOL is responsible for compensation decisions under the Energy Employees 

Occupational Illness Compensation Program Act.

U.S. Energy Research and Development Administration (ERDA)

One of the successors to the U.S. Atomic Energy Commission, the duties of which ERDA and 

the U.S. Nuclear Regulatory Commission assumed in 1974. ERDA assumed the responsibility 

for development of new reactors along with production and use of nuclear materials by the 

Federal government, and the U.S. Nuclear Regulatory Commission assumed responsibility for 

oversight of the commercial nuclear power industry. The U.S. Department of Energy 

succeeded ERDA in 1979.

X-ray radiation 

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.