ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller & Associates I MJW Corporation

Page 1 of 55

Document Title:

Idaho National Laboratory – 

Occupational Environmental Dose

Document Number: ORAUT-TKBS-0007-4 

Revision: 01

Effective Date: 08/17/2007

Type of Document TBD

Supersedes: Revision 00 

Subject Expert(s): N/A

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

Approval: Signature on File Approval Date: 08/14/2007

Jo Ann M. Jenkins, Document Owner

Concurrence: Signature on File Concurrence Date: 08/14/2007

John M. Byrne, Task 3 Manager 

Concurrence: Signature on File Concurrence Date: 08/14/2007

Edward F. Maher, Task 5 Manager 

Concurrence: Signature on File Concurrence Date: 08/15/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 08/17/2007

James W. Neton, Associate Director for Science 

 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. 01 Effective Date: 08/17/2007 Page 2 of 55

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.
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 3 of 55

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............................................................................................. 27

4.2.2.1 Stationary Low-Power Reactor No. 1 Accident ..............................................27

4.2.2.2 Criticalities ......................................................................................................27

4.2.2.2.1 INTEC Criticality of October 16, 1959........................................28

4.2.2.2.2 INTEC Criticality of January 25, 1961........................................28

4.2.2.2.3 INTEC Criticality of October 17, 1978........................................29

4.2.2.2.4 HTRE No. 3 Criticality Excursion (IET 13) ................................. 29

4.2.2.3 Releases from Planned Tests ........................................................................29

4.2.2.3.1 Fuel Element Burn Test A..........................................................30

4.2.2.3.2 Fuel Element Burn Test B..........................................................30

4.2.2.3.3 Fuel Element Cutting Facility Filter Break..................................30

4.2.2.3.4 Initial Engine Test 14 ................................................................. 32

4.2.2.3.5 Initial Engine Test 15(B) ............................................................32

4.2.2.3.6 Initial Engine Test 16 ................................................................. 33

4.2.2.3.7 Initial Engine Test 17(B) ............................................................34

4.2.2.3.8 Initial Engine Test 18 ................................................................. 34

4.2.2.3.9 Initial Engine Test 19(A) ............................................................35

4.2.2.3.10 Initial Engine Test 25(A) ............................................................35

4.2.2.3.11 Initial Engine Test 25(B) ............................................................36

4.2.2.3.12 Initial Engine Test 26(A) ............................................................36

4.3 External Dose ...........................................................................................................................36

4.3.1 Facility Fence-Line Annual Doses................................................................................ 37

4.3.2 Facility Air Immersion Doses........................................................................................ 39

4.4 Uncertainty................................................................................................................................40

4.5 Attributions and Annotations .....................................................................................................41

References ...........................................................................................................................................45
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LIST OF TABLES

TABLE TITLE

4-1 Intake by year for ANL-W, 1952 to 2005...................................................................................

PAGE

15

4-2 Intake by year for ARA, 1952 to 2005.......................................................................................17

4-3 Intake by year for CFA, 1952 to 2005.......................................................................................18

4-4 Intake by year for INTEC, 1952 to 2005 ...................................................................................19

4-5 Intake by year for RWMC, 1952 to 2004...................................................................................20

4-6 Intake by year for SPERT, 1952 to 2005 ..................................................................................21

4-7 Intake by year for TAN, 1952 to 2005.......................................................................................22

4-8 Intake by year for TRA, 1952 to 2005.......................................................................................23

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

monitoring results......................................................................................................................24

4-10 Intakes for criticalities that occurred at the INEEL ....................................................................28

4-11 Intakes for special tests at INL..................................................................................................31

4-12 Intakes for initial engine tests at INL.........................................................................................32

4-13 INL facility fence direct gamma values .....................................................................................39

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 (hr2

/m3 × 10-9), normalized to 

unit release rate ........................................................................................................................14

4-4 Onsite and distant particulate beta concentrations in air ..........................................................25

4-5 CFA air monitoring (gross beta) data for 1976 ...........................................................................26

4-6 Dispersion coefficient contours for the SL-1 accident...............................................................27

4-7 Example of onsite TLD monitoring data....................................................................................38
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 5 of 55

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

CFA Central Facilities Area

Ci curie 

cm centimeter

CPP Chemical Processing Plant 

d day

DOE U.S. Department of Energy

EBR Experimental Breeder Reactor

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) 

H&S Health and Safety (division) 

hr hour

HTRE Heat Transfer Reactor Experiment

ICPP Idaho Chemical Processing Plant

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

INTEC Idaho Nuclear Technology and Engineering Center

km kilometer

kW kilowatt

LOFT Loss-of-Fluid Test

LPTF Low Power Test Facility
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 6 of 55

m meter

M moderate (solubility rate)

mi mile

min minute

mL milliliter

mo month

mR milliroentgen

mrem millirem

MW megawatt

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

ORAU Oak Ridge Associated Universities

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

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

WWP Warm Waste Pond

yr year

β beta

γ gamma

μCi microcurie 

°F degrees Fahrenheit
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 7 of 55

§ section or sections
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 8 of 55

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

As noted above, the statute 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 contains an exclusion with respect to the Naval Nuclear Propulsion Program, 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 dosimetry monitoring results are considered valid for use in 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:

] 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) define “performance of duty” for DOE employees 

with a covered cancer or restrict the “duty” to nuclear weapons work.

• Radiation from naturally occurring radon present in conventional structures

• Radiation from diagnostic X-rays received in the treatment of work-related injuries

 

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. 01 Effective Date: 08/17/2007 Page 9 of 55

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)]. 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 facility. The releases discussed here have been divided into (1) normal 

operational releases and (2) episodic releases that generally have been of short duration. 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 fence-line doses (minus background) are presented 

for 11 locations. 

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

As outlined and discussed in the latest revision of Section 2 of this Site Profile (ORAUT 2005), 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

During the 50-yr 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, 

and (3) the Experimental Breeder Reactor No. 2 (EBR-II), at Argonne National Laboratory–West 

(ANL-W) at the southeastern 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. Throughout its history, the Chem Plant, or CPP as it is commonly known, has reprocessed 
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 10 of 55

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 will be 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. Figure 

4-1 shows the locations of INL facilities. 

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

SOUTH 

SECURITY GATE

▪

▪ ▪ CERRO GRANDE

FRENCHMAN’S

 CABIN

▪

RENO RANCH
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 11 of 55

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-yr 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 that time, 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 program (RSAC-4; Wenzel 1990). The 

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

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

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 code (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 code, see Peterson (2004).

Figure 4-2 shows the chronology of facilities and programs at the 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 it into context with the timeline in Figure 4-2.

The Experimental Breeder Reactor No. 1 (EBR-I) was the first reactor to operate at the site. It and the 

Boiling-Water Reactor Experiments (BORAX-I through -V) were in the southwestern corner of the site,
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 12 of 55

and they 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 the EBR-II, the Transient Reactor and 

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

Experiment Test (TREAT) Facility, the Zero Power Plutonium (later Physics) Reactor, 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 Reactor Experiment, 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, 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.
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 13 of 55

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 INEL 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-yr average of mesoscale dispersion isopleths of groundlevel air concentrations (DOE 1991) shown in Figure 4-3 was used.2

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 

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

(LPTF), depending on the year of operation (ORAUT 2005). The SPERT area includes the Waste 

Experimental Reduction Facility 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, 

 For 1993 to 2005, 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) were used to calculate the facility annual concentration. 

 

2 As used at INL, this quantity is the sum of 8,766 calculations of the hourly average χ/Q. 
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 14 of 55

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.

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). 

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 following discussion provides information found

in site documentation about facility environmental sampling and monitoring and provides data that can 

be compared with these calculations. 

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 [6]. The early IDO Health and Safety 

(H&S) Division annual reports (AEC 1960 -1970b) 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
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 15 of 55

(Dodd 1964) and a routine facility environmental monitoring program was developed between 1963 

and 1970. In 1968 and 1969, formal annual EMRs report 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, 

Table 4-1. Intake (Bq/yr) by year for ANL-W, 1952 to 2005 [7]. 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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 

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

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

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

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

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

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

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

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

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

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

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
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 16 of 55

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

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

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

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

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

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
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 17 of 55

Table 4-2. Intake (Bq/yr) by year for ARA, 1952 to 2005 [7]. 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 18 of 55

Table 4-3. Intake (Bq/yr) by year for CFA, 1952 to 2005 [7]. 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 19 of 55

Table 4-4. Intake (Bq/yr) by year for INTEC, 1952 to 2005 [7]. 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 20 of 55

Table 4-5. Intake (Bq/yr) by year for RWMC, 1952 to 2004 [7]. 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 21 of 55

Table 4-6. Intake (Bq/yr) by year for SPERT, 1952 to 2005 [7]. 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 22 of 55

Table 4-7. Intake (Bq/yr) by year for TAN, 1952 to 2005 [7]. 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 23 of 55

Table 4-8. Intake (Bq/yr) by year for TRA, 1952 to 2005 [7]. 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 24 of 55

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 of 5.3. Nevertheless, this 

comparison provides confidence in the results of Table 4-1. The value at the upper 84% confidence 

value 1.9 (0.36*5.3) is less than the default geometric standard deviation of 3 assumed for 

environmental results [8].

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, p14 0.18 0.01a

β 0.64 pCi/m3 AEC 1969a, p14 56 337a

I-131 <0.08 pCi/m3 AEC 1969a, p14 <7.1 1.4a

1969 α 0.003 pCi/m3 AEC 1970a, p14 0.27 2.4E-3a

β 0.501 pCi/m3 AEC 1970a, p14 44 118a

I-131 <0.014 pCi/m3 AEC 1970a, p14 <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 et al. 1987 3,290 ± 1330 890e

1988 Kr-85 at CFA 108 ± 69 pCi/m3 Hoff et al. 1989 9,770 ± 6130 14,000e

1990 Kr-85 at CFA 27.7 pCi/m3 Hoff et al. 1991 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. 

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

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-yr period has not differed from the 

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

shows the same is true for the years after 1986 (Hoff, Chew, and Rope 1987; Chew and Mitchell 
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 25 of 55

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). 

Figure 4-4. Onsite and distant particulate beta concentrations in air (Hoff, Chew, and 

Rope 1987, p. 17). 
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 26 of 55

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-mo influence on the average concentration for all the air concentrations 

measured, on and off the site. In the history of air monitoring at the INL, an operational release has 

never approached the magnitude of perturbation created by fallout as illustrated in Figure 4-5.

Figure 4-5. CFA air-monitoring (gross beta) data for 1976 (Williamson 1977a, p. 13). 
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 27 of 55

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 yr 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 that the cloud traveled from the reactor building were carefully 

monitored and well-documented (Gammill 1961; Horan and Gammill 1961; Kunze 1962). All 

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).

Figure 4-6. Dispersion coefficient contours for the SL-1 accident (redrafted from Horan and 

Gammill 1961).

4.2.2.2 Criticalities

Three accidental criticalities have occurred at the INTEC (formerly the ICPP). 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 southern end of 
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 28 of 55

INTEC and at CFA. A conservative analysis, described below, defined the potential radiological 

exposures that could have occurred to individuals at these locations. 

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 

conditions were modeled so that the χ/Q at 22 km matched the value calculated for Frenchman’s 

Cabin (south of the INL, as shown on Figure 4-1) where offsite doses were calculated and reported in 

DOE (1991). RSAC-6 was used to calculate χ/Q values for the south end of the 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 southern end of INTEC and CFA. 

Table 4-10. Intakes (Bq/event) for criticalities that occurred at the INEEL [9]. 

Date 10/16/1959 10/16/1959 1/25/1961 1/25/1961 10/17/1978 10/17/1978 11/18/1958

Event Criticality Criticality Criticality Criticality Criticality Criticality IET 13

Exposure 

location South ICPP CFA Area

South 

ICPP CFA Area South ICPP CFA Area LPTF/STPF

Rb-89 5.2E+5 8.8E+4 1.5E+1 1.0E+0 2.2E+3 3.7E+4 8.1E+2

Sr-91 5.4E+4 9.2E+3 1.7E+2 1.1E+1 1.7E+1 7.0E+2 4.1E+2

Sr-92 6.5E+4 1.1E+4 1.5E+2 9.7E+0 5.6E+0 2.1E+2 1.9E+2

Y-92 7.5E+3 1.3E+3 7.3E+1 4.9E+0 9.1E-2 2.6E+1 5.6E+1

Y-93 5.9E+3 1.0E+3 1.9E+1 1.3E+0 1.3E-1 1.5E+1 2.0E+1

Te-133 1.1E+3 8.0E+3

I-131(elem.) 2.9E+2 5.0E+1 1.9E+0 1.3E-1 3.9E-2 1.7E+0 3.4E+0

I-133 6.2E+3 1.1E+3 3.8E+1 2.5E+0 4.4E+0 5.8E+2 7.1E+1

I-134 4.6E+4 7.8E+3 7.2E+1 4.8E+0 1.1E+2 3.4E+3 2.0E+2

I-135 2.0E+4 3.4E+3 1.1E+2 7.2E+0 5.5E+1 2.2E+3 4.6E+2

Cs-138 7.7E+5 1.3E+5 5.2E+2 5.6E+4 5.0E+3

Ba-139 4.0E+5 6.8E+4 6.6E+2 4.4E+1 1.4E+2 1.8E+4 3.4E+3

La-141 2.8E+4 4.7E+3 1.0E+2 6.8E+0 4.6E-1 9.0E+1 2.0E+2

La-142 2.3E+4 3.9E+3 4.3E+1 2.8E+0 4.8E-1 5.7E+1 9.3E+1

U-234 7.1E-2

Dose fract 93.9% 92.5% 94.5% 94.5% 98.3% 97.1% 94.4%

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 states:

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)
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 29 of 55

As for the 1959 criticality, Χ/Q values were calculated for the southern end of INTEC and CFA. 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 southern end of INTEC and CFA 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 min 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. 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 

southern 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 (Linn 1962) about the 

test, 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 
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 30 of 55

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, or CFA. 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.

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 hr and decayed for 70 d, 

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 hr using jet fuel as the burning agent. 

After 2 hr, 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 d 

(Brodsky and Beard 1960, p. 37). 

After 90 s, most of the fuel element had melted and dispersed. DOE (1991, Figure 3-12) shows the 

trajectory of the plume, and Figure 3-13 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 Fuel Element Cutting Facility 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. 
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 31 of 55

Table 4-11. Intakes (Bq/event) for special tests at INL [12]. 

Date Event

Exposure 

location Sr-89 Sr-90 Y-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
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 32 of 55

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. 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-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 hr 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 hr 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 the IET 14 test and the intakes at TRA/INTEC

and CFA. 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 doses are for the entire

exposure period of 26 d.

Table 4-12. Intakes (Bq/event) for initial engine tests at INL [13]. 

Period Test

Exposure 

location Rb-89 Sr-89

I-131 

(elem.) I-133 I-135 Cs-138 U-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

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

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

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

1/26–2/7/60 IET 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

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

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

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:
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 33 of 55

• Endurance capabilities of the advanced metals at a design temperature of 2,000°F for 

extended periods (planned endurance testing to total 120 hr 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 hr at an insert extrapolated fuel sheet 

temperature of 2,015o

F. The operation was terminated after 80.75 hr 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. 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 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 and CFA. An individual would 

only have been exposed to these concentrations and intakes if present at the locations between June 

16 and June 24, 1959. The radiological doses in Table 4-12 are for the entire 9-d 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, 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 hr), 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.
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 34 of 55

4.2.2.3.7 Initial Engine Test 17(B) 

IET 17 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 doses in Table 4-12 are 

for the entire exposure period of 40 d.

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 

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. 
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 35 of 55

4.2.2.3.9 Initial Engine Test 19(A) 

IET 19, 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 hr 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-min-decayed curies. The total fission product release reported for 

IET 19 was 2,892 Ci. The release for this test was modeled like that for DOE (1991). Table 4-12 lists 

intakes for TRA/INTEC 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 

doses in Table 4-12 are for the entire exposure period of 13 d.

4.2.2.3.10 Initial Engine Test 25(A) 

IET 25, 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 

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 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 
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 36 of 55

November 22 and December 15, 1960. The tabulated intakes are for the entire exposure period of 

24 d.

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 the IET 25 test (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 d. 

4.2.2.3.12 Initial Engine Test 26(A) 

IET 26, 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 and TRA/INTEC and CFA. The intakes are listed in Table 4-12 for 

TRA/INTEC 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 d. 

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. 

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 fence-line 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).
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 37 of 55

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 southern 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 mo 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 the 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 Fence-Line Annual Doses

Between the latter part of 1970 and the latter part of 1972, facility fence-line 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. 

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 fence-line TLD measurements, made on a 6-mo 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, 1980 b,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-mo period. For this particular 2-yr 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 assigned a 

2-sigma uncertainty of 11% to 15%. However, 80% of the 136 values ascribed for the 34 locations 

over the 2-yr period have a 2-sigma uncertainty of less than 10% [17]. 

To supply facility values for the period from 1952 to 1972, the highest 6-mo value from April 1972 to 

April 1973 for a facility was multiplied by 2 and applied to each year between 1952 and 1972 [18] (Not 
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 38 of 55

Figure 4-7. Example of onsite TLD monitoring data (Williamson 1976a, p. 38). 

all the listed facilities commenced operation or even existed in 1952.). Facility fence-line 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 

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 

fence-line TLD measuring periods were recorded in the EMRs. All reduced facility fence-line TLD 

data (facility fence-line data minus background) in the EMRs are listed in Table 4-13 (AEC 1971, 
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 39 of 55

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). A more detailed discussion of the data 

is included in Peterson (2004).

Table 4-13. INL facility fence direct gamma values (TLD – background) (mR) [19]. 

Year

ARA 

I & II SPERT

TANTSFTANLOFTTANLPTF CFA TRA INTEC RWMC EBR-II TREATBackground

TLDs used

1, 2, 

3 All All All All All

1, 7, 

12, 

13

2, 11, 

12, 13

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

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 fence-line TLD doses listed in Table 4-13.

Facility Air Immersion Doses
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 40 of 55

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 IET tests 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]. 

During 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 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. 
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 41 of 55

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-mo 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 fence-line 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 INTEC 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/or 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 upon 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/or 

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, 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.
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 42 of 55

[3] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

Figure 4-3 shows a concentration isopleth of about 5 for Idaho Falls compared to 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.

An example of such analysis is Ebersole (1956).

[7] 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 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.

[8] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

The default geometric standard deviation of 3 has been confirmed by phone conversation and 

e-mail from C. Bloom to N. Rohrig of 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.

[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.
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 43 of 55

[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.

This comes from counting the values in Figure 4-7.

[18] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

This explains how missing data were generated.

[19] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

See the references listed in the last paragraph of Section 4.3.1.

[20] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

Discussion between Henry Peterson, Doug Wenzel, and Richard Dickson in June 2003.

[21] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

These results are shown in the environmental monitoring reports, for example Walker (1972).

[22] Peterson, Henry. ORAU Team. Site Expert. September-December 2003.

Discussion between Henry Peterson, Doug Wenzel, and Richard Dickson in June 2003.

[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.
Document No. ORAUT-TKBS-0007-4 Revision No. 01 Effective Date: 08/17/2007 Page 44 of 55

[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. 01 Effective Date: 08/17/2007 Page 45 of 55

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