ORAU TEAM 

Dose Reconstruction 

Project for NIOSH

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

Page 1 of 39

Document Title: 

Argonne National Laboratory – West – 

Occupational Environmental Dose

Document Number: ORAUT-TKBS-0026-4 

Revision: 01 

Effective Date: 06/05/2007 

Type of Document TBD 

Supersedes: Revision 00 

Subject Expert(s): William E. Kennedy and Henry Peterson 

Site Expert(s): Norman D. Rohrig 

Approval: Signature on File Approval Date: 05/24/2007 

Jo Ann M. Jenkins, Document Owner 

Concurrence: Signature on File Concurrence Date: 05/23/2007 

John M. Byrne, Task 3 Manager 

Concurrence: Signature on File Concurrence Date: 05/23/2007

Edward F. Maher, Task 5 Manager 

Concurrence: Signature on File Concurrence Date: 05/31/2007

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 06/05/2007

 Brant A. Ulsh, PhD CHP 

 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-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 2 of 39 

PUBLICATION RECORD 

EFFECTIVE 

DATE 

REVISION 

NUMBER DESCRIPTION 

09/30/2004 00 New Technical Basis Document for the Argonne National Laboratory 

– Occupational Environmental Dose. Incorporates NIOSH and 

internal review comments. First approved issue. Initiated by Norman 

D. Rohrig. 

06/05/2007 01 Revised for Biennial review. Incorporates Attribution and Annotation 

section. Constitutes a total rewrite of the document. Incorporates 

formal internal and NIOSH review comments. This revision results in 

no change to the assigned dose and no PER is required. Training 

required: As determined by the Task Manager. Initiated by Jo Ann 

M. Jenkins. 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 3 of 39 

TABLE OF CONTENTS 

SECTION TITLE PAGE

Acronyms and Abbreviations..................................................................................................................5

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

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

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

4.2 Internal Intakes from Onsite Airborne Radionuclide Concentration ..........................................12

4.2.1 Operational Releases................................................................................................... 12

4.2.2 Episodic Releases at INL............................................................................................. 16

4.2.2.1 Stationary Low-Power Reactor No. 1 Accident ..............................................18

4.2.2.2 Criticality and Accident Occurrences at Idaho Chemical 

Processing Plant ............................................................................................18

4.2.2.2.1 Criticality of October 16, 1959 ......................................................19

4.2.2.2.2 Criticality of January 25, 1961 ......................................................19

4.2.2.2.3 Fuel Element Cutting Facility Filter Break.....................................20

4.2.2.3 Releases from Planned Tests ........................................................................20

4.2.2.3.1 Initial Engine Test 14 ....................................................................20

4.2.2.3.2 Initial Engine Test 15(B) ...............................................................21

4.2.2.3.3 Initial Engine Test 17(B) ...............................................................21

4.2.2.3.4 Initial Engine Test 19(A) ...............................................................22

4.2.2.3.5 Initial Engine Test 25(A) ...............................................................22

4.2.2.3.6 Initial Engine Test 26(A) ...............................................................23

4.3 External Dose ...........................................................................................................................23

4.3.1 Facility Fenceline Annual Doses.................................................................................. 23

4.3.2 Facility Air Immersion Doses........................................................................................ 25

4.4 Uncertainty................................................................................................................................25

4.4.1 Operational Releases................................................................................................... 26

4.4.2 Episodic Releases........................................................................................................ 26

4.4.3 Film Badge and Thermoluminescent Dosimeter Measurements ................................. 26

4.5 Attributions and Annotations.....................................................................................................26

References ...........................................................................................................................................31
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 4 of 39 

LIST OF TABLES 

TABLE TITLE PAGE

4-1 Intake by year for ANL-W, 1952–2005......................................................................................14

4-2 Comparison of calculated facility intakes with intakes from environmental 

monitoring results......................................................................................................................15

4-3 Intakes at EBR-I for criticalities at ICPP....................................................................................19

4-4 Intakes at EBR-I for FECF Filter Break incident at ICPP ..........................................................20

4-5 Intakes at EBR-I for initial engine tests at INL ..........................................................................21

4-6 ANL-W facility fenceline direct gamma values (TLD minus background) .................................25

LIST OF FIGURES 

FIGURE TITLE PAGE

4-1 INL site map..............................................................................................................................10

4-2 Nine-year average mesoscale dispersion isopleths of air concentration at 

ground level normalized to unit release rate.............................................................................13

4-3 Onsite and distant particulate beta concentrations in air ..........................................................17

4-4 Dispersion coefficient contours for SL-1 accident.....................................................................18

4-5 Example of onsite TLD monitoring data....................................................................................24
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 5 of 39 

ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission 

AMAD activity median aerodynamic diameter 

ANL-W Argonne National Laboratory–West 

ARA Auxiliary Reactor Area 

ATR Advanced (previously Army) Test Reactor 

BORAX Boiling Water Reactor Experiment 

Bq becquerel 

CFA Central Facilities Area 

CFSGF Coal-Fired Steam Generating Facility 

CFR Code of Federal Regulations 

Ci curie 

CTF Core Test Facility 

d day 

DOE U.S. Department of Energy 

EBR Experimental Breeder Reactor 

ECF Expended Core Facility 

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000 

EFS Experimental Field Station 

EMR Environmental Monitoring Report 

EOCR Experimental Organic-Cooled Reactor 

ETR Engineering Test Reactor 

F fast (solubility rate) 

FAST Fluorinel Dissolution Process and Fuel Storage (facility) 

fCi femtocurie 

FECF Fuel Element Cutting Facility 

ft foot 

GE-ANP General Electric-Advanced Nuclear Propulsion (Program) 

H&S Health & Safety 

hr hour 

HTRE Heat Transfer Reactor Experiment 

ICPP Idaho Chemical Processing Plant 

IDO Idaho Operations Office 

IET Initial Engine Test 

INEEL Idaho National Engineering and Environmental Laboratory 

INEL Idaho National Engineering Laboratory 

INELHDE Idaho National Engineering Laboratory Historical Dose Evaluation

INL Idaho National Laboratory 

INTEC Idaho Nuclear Technology and Engineering Center 

km kilometer 

kW kilowatt 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 6 of 39 

m meter 

M moderate (solubility rate) 

mi mile 

mL milliliter 

mR milliroentgen 

mrem millirem 

MTR Materials Test Reactor 

NRF Naval Reactors Facility 

NCRP National Council on Radiation Protection and Measurements 

NIOSH National Institute for Occupational Safety and Health 

NOAA National Oceanic and Atmospheric Administration 

NRTS National Reactor Testing Station 

NWCF New Waste Calcining Facility 

ORAU Oak Ridge Associated Universities 

pCi picocurie 

POC probability of causation 

PREPP Process Experimental Pilot Plant 

RM radioactive material 

RSAC Radiological Safety Analysis Computer (program) 

RWMC Radioactive Waste Management Complex 

S slow (solubility rate) 

SL-1 Stationary Low-Power Reactor No. 1 

SPERT Special Power Excursion Reactor Test (facility) 

SRDB Ref ID Site Research Database Reference Identification 

STSL Semi Scale Support Laboratory 

SWEPP Stored Waste Examination Pilot Plant 

TAN Test Area North 

TBD technical basis document 

TLD thermoluminescent dosimeter 

TRA Test Reactor Area 

TREAT Transient Reactor Test 

TSF Technical Support Facility 

U.S.C. United States Code 

WERF Waste Experimental Production Facility 

wk week 

WRRTF Water Reactor Research Test Facility 

yr year 

ZPPR Zero Power Plutonium (later Physics) Reactor 

α alpha activity 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 7 of 39 

β beta activity 

γ gamma activity 

μCi microcurie 

°F degree Fahrenheit 

§ section or sections 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 8 of 39 

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

] 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. 

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: 

• 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-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 9 of 39 

4.1.1 Purpose

This TBD provides the basis for the environmental doses at the Argonne National Laboratory–West 

(ANL-W) facility at the Idaho National Laboratory (INL). This information can be used in dose 

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

This TBD addresses radioactive material (RM) releases from areas or facilities at INL, formerly the 

National Reactor Testing Station (NRTS) and then the Idaho National Engineering Laboratory (INEL) 

and the Idaho National Engineering and Environmental Laboratory (INEEL), that could affect 

employees at the ANL-W facility. The analysis for this TBD divided releases into two components: 

(1) normal “chronic” operational releases, and (2) episodic releases that generally are of short 

duration (DOE 1991). These releases potentially represent unrecorded or missed doses, as either 

direct gamma radiation or beta-gamma radiation from immersion in the radioactive gaseous cloud, for 

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

inhalation. 

In addition, this TBD 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, 

these facility background doses were recorded by film badges infrequently and inconsistently before 

1970 and by thermoluminescent dosimeters (TLDs) on a routine basis since 1970 (Walker 1971). 

These facility background doses, or facility fenceline doses, as they are sometimes called, are a 

nebulous indication of a dose that workers could receive if they inhabited outside areas within the 

facility. This TBD presents ANL-W facility fenceline doses (minus background) for the Environmental 

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

facility and EBR-II locations, both for 1972 to the present. 

4.1.2 Scope

As outlined and discussed in Section 2.0 of the ANL-W site profile (ORAUT 2006), the U.S. Atomic 

Energy Commission (AEC) selected the INL site as an isolated location for the testing of reactor 

concepts. INL is isolated from the public in two important aspects: (1) it is remote from population 

centers, and (2) it is isolated hydrologically because no surface streams originate on the site 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 5 to 15 ft/d (ERDA 1977a). 

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

of the DOE Chicago Operations Office. Although the facility operates in accordance with 10 C.F.R. pt. 

835, its operations and atmosphere have been and are more in line with a university engaged with 

pure nuclear energy research; these operations support those of the University of Chicago and 

Argonne National Laboratory–East, near Chicago, Illinois. During its first 14 years of operation (i.e., 

from 1951 through 1965), ANL-W was in the southwest corner of INL at the location of EBR-I. At this 

location, ANL-W conducted tests on EBR-I, Zero Power Reactor No. 3, Argonne Fast Source Reactor, 

and all Boiling Water Reactor Experiments (BORAX). In 1958, construction began on the TREAT 

Facility and on EBR-II at the present ANL-W location at the southeastern corner of INL. Since the 

mid-1970s, essentially all ANL-W operations have been conducted at the present ANL-W location, as 

shown in Figure 4-1. 

During the 50-year history of the INL site, DOE and its predecessor agencies designed, built, and 

operated about 50 different conceptual reactors. All of these reactors have been prototype, lowpower critical, or test reactors. INL operated no weapons production or commercial power reactors. 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 10 of 39 

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

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 INL 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) EBR-II at ANL-W at the southeastern corner of the site, which 

has produced minor amounts of airborne effluent (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, which was constructed in the early 1950s, began processing nuclear fuel in February 1953 

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

it is commonly known, has reprocessed fuel from test reactors at INL, zirconium-clad fuel reclaimed 

from various reactors, stainless-steel-clad fuel from EBR-II, and fuel from many AEC test reactors 

around the world. Apart from the GE-ANP Program, which tested nuclear-powered aircraft engine 

concepts with only one barrier (fuel cladding) between the fission products and the environment, TRA 

and ICPP airborne releases have been the most radiologically significant releases at INL (DOE 1991; 

SOUTH 

SECURITY GATE

▪

▪ ▪ CERRO GRANDE

FRENCHMAN’S

CABIN

▪

RENO RANCH
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 11 of 39 

Till et al. 2002). Through the years that Environmental Monitoring Reports (EMRs) have been 

published, INTEC airborne effluents have been attributed to creation of the majority of the INL 

boundary dose (Till et al. 2002). Therefore, it should be suspected that INTEC airborne effluent would 

also be responsible for the largest INL worker doses. Calculations performed for this TBD show that 

although ICPP airborne effluent is the most radiologically significant release at INL, the impact to all 

facility workers is significantly below the allowable and acceptable limits (Peterson 2004). Figure 4-1 

shows INL facility locations, including the EBR-I and ANL-W facilities. 

From the beginning of operations at the INL site, DOE and its predecessor agencies selected facility 

locations 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 its general meteorological characterization 

indicates a nocturnal inversion from the north-northeast and a daytime lapse condition with winds from 

the southwest, transitional weather regimes are less frequent than at lower elevations. The 50-year 

history of the site has demonstrated that the large expanse of INL and this meteorological 

characteristic have been satisfactorily effective in maintaining the operational isolation of each facility 

[1]. 

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

capability of plume tracking by aircraft. The local National Oceanic and Atmospheric Administration 

(NOAA) field office, which was dedicated to INL 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 from the GE-ANP tests, the 

Controlled Environmental Release Test, Fission Product Field Release Test, Fuel Element Burn Tests 

A and B, etc., were directed over an instrumented monitoring grid (GRID III) that was remote from 

other facilities, such that releases did not affect other onsite facilities [2]. 

INL reviewed and analyzed all airborne releases that have occurred since the beginning of site 

operations as a result of a request from the DOE Idaho Operations Office (IDO) to evaluate the 

radiological impact to individuals at the INL boundary from airborne releases that had occurred since 

the beginning of site operations (DOE 1991). With the help of NOAA, which had hourly 

meteorological data from 1956 to that time, INL completed analyses for all airborne releases that 

occurred at the site. Radiological consequences for an adult, a child, and an infant were calculated 

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

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

Evaluation (DOE 1991; this TBD refers to that report as the INELHDE). All releases considered for 

that report are the bases for the releases considered in this TBD. In addition, all releases 

documented in the INELHDE, operational and episodic, have been independently reviewed and 

found, with minor modifications, to be substantially appropriate. The review, which was conducted by 

Radiological Assessment Corporation at the request of the Centers for Disease Control and 

Prevention and the State of Idaho, evaluated the methodology by which the RSAC-4 program 

performs dose calculations against the methodology favored by the National Council on Radiological 

Protection and Measurements (NCRP). 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 program (Wenzel and Schrader 2001) is used extensively in this document to 

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

information on the RSAC program, see Peterson (2004). 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 12 of 39 

EBR-I was the first reactor to operate at the site. It and BORAX-I through -V, which were in the 

southwestern corner of INL, were operated for the AEC Chicago Operations Office by the University 

of Chicago as ANL-W. These low-power reactors produced essentially no 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 INL site where the locations of EBR-II, the 

TREAT Facility, the Zero Power Plutonium Reactor, etc., are indicated in Figure 4-1. The EBR-I 

location is now a historic landmark and is open to the public during the summer months. 

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

the ANL-W facility – at the southwest location through 1965 and at the southeast location beginning in 

1966 [3]. 

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. 

4.2 INTERNAL INTAKES FROM ONSITE AIRBORNE RADIONUCLIDE CONCENTRATION 

This section addresses onsite concentrations of radionuclides and onsite internal personnel intakes 

from normal operational releases and from shorter term releases such as those from criticality 

incidents at ICPP and tests performed by the GE-ANP Program at TAN. As stated above, operational 

releases from ICPP and TRA have been the predominant and radiologically significant releases at INL 

during the history of the site (Till et al. 2002). For more discussion of these releases and their 

relationship to less significant releases, see Peterson (2004), DOE (1991), or Till et al. (2002). 

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

consideration. When solubility is of concern, an insoluble oxide form for solids is recommended for 

analysis, with type S and M materials being the predominant forms [4]. Without more definitive 

information on the type of material, the dose reconstructor should use the material that maximizes the 

dose for a particular situation. When iodines are of concern, the dose reconstructor should consider 

them to be a vapor intake of type F materials [5]. 

4.2.1 Operational Releases

Estimation of onsite concentrations of radionuclides and resulting potential intakes from operational 

releases at INL facility locations employs the same methodology used to determine offsite 

concentrations for site annual EMRs. The release for each year of operation is exactly the same as 

that documented in the INELHDE (DOE 1991) with one exception: An analysis to reduce the number 

of radionuclides and yet retain those that contributed about 95% of the inhalation dose (see Peterson 

2004) reduced the number of radionuclides from 56 to 9 for the operational releases. 

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

mesoscale dispersion isopleths of ground-level air concentrations in the EMRs, as described in the 

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

available for years beginning in 1973, with the exception of 1978, when INL upgraded the telemetry 

system. For years before 1973, the TBD analysis used a 9-year average of mesoscale dispersion 

isopleths of ground-level air concentrations (DOE 1991); see Figure 4-2. For 1993 to 2002, the 

analysis used annual average mesoscale isopleths from the annual environmental reports (Mitchell 

1994, 1995, 1996, 1997; Evans et. al. 1998,; Saffel, et. al. 2000; Stoller 2002a,b,c, 2004, 2005) to 

calculate the facility annual concentration. 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 13 of 39 

Of the many facilities on INL, this analysis considered only the ANL-W facility. Yearly isopleth values 

for each ANL-W location (EBR-I for 1952 to 1965 and EBR-II for 1966 to the present) have been 

Figure 4-2. 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). 

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. Table 4-1 lists these values. 

The annual inhaled quantities (becquerels per year) in Table 4-1 are based on known and published 

INL annual airborne emissions. The following discussion provides information from NRTS/INEL/ 

INEEL/INL documentation on facility environmental sampling and monitoring and on data that can be 

 

2

 As used at the INL, this quantity is the sum of 8,766 calculations of the hourly average χ/Q. 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 14 of 39 

compared with the calculated inhaled quantities of Table 4-1. Environmental air sampling at the 

facility areas has been performed at least since the mid-1950s where airborne effluents were known 

or suspected to exist [7]. The early IDO Health and Safety (H&S) Division Annual Reports document 

Table 4-1. Intake (Bq/yr) by year for ANL-W, 1952 to 2005 [6]. 

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 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 

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 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 15 of 39 

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 

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 did not occur in a set 

location or for a standard duration. The 1963 Annual Progress Report of the IDO H&S Division (AEC 

1964) contains some facility environmental monitoring data. INL developed a routine facility 

environmental monitoring program between 1963 and 1970. In 1968 and 1969, formal EMRs (AEC 

1968, 1969a,1969b, 1970a) reported alpha, beta, and 131I concentrations that can be correlated with 

the data of Table 4-1. The 1970 EMR (AEC 1970b, 1971) reports gross beta values measured at the 

Central Facilities Area (CFA) that can be correlated with Table 4-1 (EBR-I) values. The analysis for 

this TBD reviewed EMRs between 1970 and 1990 for data that are available and usable for this 

correlation. 

Table 4-2 lists results of the comparison. Because of the large variation in measurements, the ratio of 

values calculated from the EMRs (Table 4-2, column 5) to those derived from the releases in Table 4-

1 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 4-2. 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) 

β-γ 17 pCi/m3 1,510 1,310a 1963 

Pu-239 0.6 fCi/m3

 0.05 0.014a

α 2.2 fCi/m3 AEC 1969a, p. 14 0.18 0.01a

β 0.64 pCi/m3 AEC 1969a, p. 14 56 337a

1968 

I-131 <0.08 pCi/m3

 AEC 1969a, p. 14 <7.1 1.4a

α 0.003 pCi/m3 AEC 1970a, p. 14 0.27 2.4E-3a

β 0.501 pCi/m3 AEC 1970a, p. 14 44 118a

1969 

I-131 <0.014 pCi/m3

 AEC 1970a, p. 14 <1.2 2.1a

Gross β 0.6 pCi/m3 AEC 1971, p. 5 53 74a 1970 

Max. gr. β at CFA 0.81 pCi/m3

 AEC 1971, p. 5 72 74a

1973 Gross β 95 +42 fCi/m3 AEC 1974, p. 14 8.4 ±3.7 0.8a

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 0.6–0.5 --- 

EBR-I 

Ce-144 4–8 fCi/m3 AEC 1974, p. 16 0.36–0.71 0.057b

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 --- 

EFS 

Cs-137 17–27 fCi/m3

 AEC 1974, p. 16 1.5–2.4 --- 

1976c Gross β 30–60 fCi/m3 ERDA 1977b 2.5– 5.3 0.6-25d

1986 Kr-85 at CFA 37 ±15 pCi/m3 Hoff, Chew, and Rope 

1987 

3,290 ±1,330 890e

1988 Kr-85 at CFA 108 ±69 pCi/m3 Hoff, Mitchell, and 

Moore 1989 

9,770 ±6,130 14,000e

1990 Kr-85 at CFA 27.7 pCi/m3 Hoff et al. 1991 2,400 690e

a. Values from the INEEL Environmental Occupational Dose TBD (ORAUT 2004, Table 4-3) for CFA. 

b. Values from ORAUT (2004, Table 4-5) for the Radioactive Waste Management Complex (RWMC) because 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. 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 16 of 39 

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 INL-released radionuclides. 

Figure 4-3 shows the variation of INL environmental sampling results for 1978 through 1986, which is 

typical for earlier and for later years (Hoff, Chew, and Rope 1987). This figure also shows the close 

correlation of environmental sample results acquired at “distant communities” and those acquired at 

INL facilities and 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 has not 

differed from distant community concentrations by more than a factor of 2 for the 9-year period and is 

very similar for earlier and later periods; subsequent EMRs show the same correlation for the years 

before 1978 and after 1986. Nearly all of the greater perturbations in facility and distant community 

concentrations correlate with fallout from nuclear tests [9]. There was no discernible evidence that 

facility effluent or resuspension affected facility concentrations. 

4.2.2 Episodic Releases at INL

Of the 108 episodic releases (accidents and planned tests) analyzed in DOE (1991), only 16 had the 

potential to affect other INL facilities [10]. Only 9 of the 16 events could have affected the EBR-I 

facility [11]. Section 4.1.2.2 describes three of these events, two of the three criticalities at ICPP, and 

the Fuel Element Cutting Facility (FECF) Filter Break, which also occurred at ICPP. The other six 

were planned GE-ANP tests that could have affected the EBR-I facility, as follows: 

1. Initial Engine Test (IET) 14 4. IET 19(A) 

2. IET 15(B) 5. IET 25(A) 

3. IET 17(B) 6. IET 26(A) 

For a given test, if there was an onsite facility between the point at which the test occurred and the 

affected offsite location, that test was conservatively assumed to have affected one or more onsite 

facility. For example, the FECF Filter Break occurred at ICPP and clearly contaminated an area south 

of ICPP (AEC 1960). According to the meteorological dispersion at the time of the filter break, the 

affected offsite location was Frenchman’s Cabin. Because EBR-I is in the straight-line path between 

ICPP and Frenchman’s Cabin, a radiological impact analysis was conducted for EBR-I [12]. 

Because all other test releases listed above, which originated at the TAN facility, affected one location 

on the southern boundary [Frenchman’s Cabin (shown on Figure 4-1), as evaluated in DOE (1991)], 

they have been assumed to have affected EBR-I because that facility is on the plume trajectory from 

TAN [13]. The following sections discuss these events. The Stationary Low-Power Reactor No. 1 

(SL-1) accident, which was widely publicized, is included only to show it did not affect any other facility 

at INL. 

An effort has been made to reduce the number of radionuclides in the release analysis for the 

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 [14]. When viewed together, the 

episodic events fall into three categories: criticalities that involved fresh fission products that have 

relatively short half-life in comparison to radionuclides released from the Fuel Element Burn Tests, for 

example; releases that involved long half-life, aged fission products (Fuel Element Burn Tests and the 

release from the FECF Filter Break); and releases from the remaining IETs that released short half-life 

radionuclides, which are generally characterized as fresh fission products, and long half-life 

radionuclides, which are characterized as aged fission products [15]. The latter category is unique to 

the GE-ANP Program because of the “direct-to-air” conversion nature of the tests [16]. Therefore, 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 17 of 39 

within these categories, the number of radionuclides has been reduced to the number that preserves 

95% of the original dose that was calculated for that particular location [17]. 

Figure 4-3. Onsite and distant particulate beta concentrations in air (Hoff, Chew, and Rope 

1987, p. 17). 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 18 of 39 

4.2.2.1 Stationary Low-Power Reactor No. 1 Accident 

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

environment. On January 3, 1961, a steam explosion at the SL-1 facility [near Auxiliary Reactor Area 

II in Figure 4-1] killed three SL-1 personnel and ruptured the SL-1 reactor vessel. This, in turn, 

propelled RM 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 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 RM. The effluent traveled to 

the south of the facility, as shown in Figure 4-4 (DOE 1991). 

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

1961). 

4.2.2.2 Criticality and Accident Occurrences at Idaho Chemical Processing Plant 

Three accidental criticalities have occurred at the ICPP (now INTEC). The first occurred on 

October 16, 1959; the second on January 25, 1961; and the third on October 17, 1978, as described 

in Sections 4.2.2.2.1, 4.2.2.2.2, and 4.2.2.2.3, respectively. The three criticalities released RM during 

or shortly after the event. In all cases, the effluent was transported to the south-southwest and 

potentially exposed personnel at the EBR-I facility because that facility was postulated to be in the 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 19 of 39 

plume path. The 1978 criticality, which released essentially just noble fission gases that were 

produced during the criticality, occurred after ANL-W operations had moved to the present southeast 

site location and, thus, did not affect personnel at the new ANL-W location. The FECF Filter Break 

accident, which also occurred at the ICPP, has been postulated to have affected the EBR-I facility and 

to have been in the path of the release plume. Analyses that were favorable to claimants, described 

below, defined the amount of potential radiological exposure that could have occurred to an individual 

at this location. 

4.2.2.2.1 Criticality of October 16, 1959 

On October 16, 1959, at approximately 3:00 a.m., a criticality event occurred at the ICPP 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) provides a full 

account of the incident and documents the calculated internal and measured external radiological 

doses for plant personnel who were involved in the incident. 

For the calculation of intakes for this incident, meteorological conditions were modeled so the Χ/Q at 

22 kilometers matched the value calculated for Frenchman’s Cabin (south of INL, as shown on 

Figure 4-1) where offsite doses were calculated and reported in DOE (1991). RSAC-6 was used to 

calculate Χ/Q values for EBR-I. 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 4-3 lists 

intakes applicable at EBR-I. 

Table 4-3. Intakes (Bq/event) at EBR-I for 

criticalities at ICPP [18]. 

Date 10/16/1959 1/25/1961 

Event Criticality Criticality 

Exposure location EBR-I Area EBR-I Area 

Rb-89 2.1E+4 4.4E-1 

Sr-91 2.2E+3 4.8E+0 

Sr-92 2.6E+3 4.3E+0 

Y-92 3.1E+2 2.2E+0 

Y-93 2.4E+2 5.7E-1 

I-131(elem.) 1.2E+1 5.7E-2 

I-133 2.6E+2 1.1E+0 

I-134 1.9E+3 2.1E+0 

I-135 8.2E+2 3.2E+0 

Cs-138 3.1E+4 

Ba-139 1.6E+4 1.9E+1 

La-141 1.1E+3 3.0E+0 

La-142 9.4E+2 1.2E+0 

4.2.2.2.2 Criticality of January 25, 1961 

The January 25, 1961, criticality occurred in ICPP vessel H-110 about 9:50 a.m. This event consisted 

of an estimated 6.0 × 1017 fissions. The report on 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 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 20 of 39 

extensive shielding provided by the process cell in which the event took place and the 

control of the fission gases by the equipment. (Paulus et al 1961)

As for the 1959 criticality, Χ/Q values were calculated for EBR-I. The source term used for this event 

is the same as that used for DOE (1991). Table 4-3 lists the intakes applicable at EBR-I if the 

individual was in the area on January 25, 1961. 

4.2.2.2.3 Fuel Element Cutting Facility Filter Break 

The ends of fuel elements sent to ICPP contained structural components 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 CPP-601 for processing occurred in the FECF in CPP-603. During the night 

of October 29 and early in the morning of October 30, 1958, INL conducted decontamination 

operations 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 ICPP (AEC 1959). 

Approximately 100 Ci of long half-life particulate RM were released over approximately 200 acres 

(AEC 1959). The released RM and quantities were the same as those published in DOE (1991). 

Table 4-4 lists the best-estimate intakes of the radionuclides in the EBR-I area. 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). 

Table 4-4. Intakes (Bq/event) at EBR-I for FECF Filter Break incident at ICPP [19]. 

Date Event 

Exposure 

location Sr-89 Sr-90 Y-91 Zr-95 Ru-103 Ru-106 

I-131 

(elem.) Ce-144 Pr-143 

10/29/58 – 

10/30/58 

FECF Filter 

Break 

EBR-I Area 6.3E-1 6.3E-1 1.4E+0 2.0E+0 1.3E-1 6.6E-1 4.2E-12 9.0E+0 2.6E-6 

4.2.2.3 Releases from Planned Tests 

The following were all planned tests under the GE-ANP Program at TAN that potentially affected the 

EBR-I facility. 

4.2.2.3.1 Initial Engine Test 14 

IET 14 was the eighth nuclear test that was 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 (Pincock 1959). 

A total of 100.25 hours was accumulated on the insert fuel cartridge at a maximum insert fuel 

temperature of approximately 2,500°F. The objectives of the test were to (1) evaluate the operational 

effect of water vapor corrosion on fueled beryllium-oxygen tubes that operated at a constant reactor 

mixed mean discharge air temperature over approximately 100 hours, and (2) measure the fission 

product release rate from uncoated fueled tubes as a function of temperature and operating time 

(Pincock 1959). 

Table 4-5 lists the fission products that were released during the IET 14 test and the intakes at EBR-I. 

An individual would have been exposed to these concentrations only if present at these locations 

between April 24 and May 19, 1959. These intakes are for a total exposure period of 26 days. 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 21 of 39 

Table 4-5. Intakes (Bq/event) at EBR-I for initial engine tests at INL [20]. 

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 EBR-I 5.8E+0 1.7E-2 2.1E+0 1.3E+1 1.8E+1 4.2E+1 3.6E-6 

6/16–6/24/59 IET 15(B) EBR-I 3.3E-1 8.4E-4 4.5E-1 2.0E+0 3.2E+0 2.3E+0 3.9E-5 

10/12–12/12/59 IET 17(B) EBR-I 2.2E-3 6.2E-3 5.7E-1 2.4E+0 2.5E+0 6.2E-1 5.5E-6 

2/17–2/29/60 IET 19(A) EBR-I 1.3E-1 6.8E-3 9.9E-1 6.7E+0 9.9E+0 4.9E+0 4.2E-7 

11/22–11/30/60 IET 25(A) EBR-I 7.1E-4 4.6E-4 3.6E-1 4.0E+0 5.1E+0 1.5E-1 1.5E-6 

12/23–12/28/60 IET 26(A) EBR-I 1.7E+0 1.1E-2 1.8E+0 7.2E+0 1.1E+1 1.5E+1 2.5E-5 

4.2.2.3.2 Initial Engine Test 15(B) 

IET 15(B) 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). From this operation, data were obtained to evaluate: 

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

extended periods (planned endurance testing to total 120 hours or more) 

2. The structural and metallurgical integrity of the fuel sheet in this particular cartridge design 

3. The nature and extent of fuel sheet damage, if any, and the effect on cartridge performance 

4. The performance potential of the cartridge 

The operation was successfully conducted to accumulate 80.75 hours at an insert extrapolated fuel 

sheet temperature of 2,015°F. The operation was terminated after 80.75 hours due to a release of 

fission products of such a quantity as 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; however, there were blisters on the inner fuel sheets. In some instances 

the blisters had ruptured. The fission product release for this test was divided into two periods based 

on a review of effluent monitoring data. The period from June 3 to 15, 1959, was considered to be an 

operation before the development of significant fuel sheet blisters. The second period was from 

June 16 to 24, 1959, when effects of blistering were clearly observed. 

According to the meteorology of the testing period, the second period affected Frenchman’s Cabin. 

Therefore, this analysis addressed the radiological impact on the EBR-I. Table 4-5 lists the release of 

fission products, which corresponds with the Part B operation release from DOE (1991) for the intakes 

applicable at EBR-I. An individual would have been exposed to these concentrations and intakes only 

if present at the locations between June 16 and June 24, 1959. The intakes are for the total 9-day 

exposure period. 

4.2.2.3.3 Initial Engine Test 17(B) 

IET 17(B) occurred between October 12 and December 12, 1959. Releases of airborne radioactivity 

occurred between November 2 and December 12, 1959, when the reactor operated at power levels 

that exceeded 100 kW. The test series involved the evaluation of the L2E-1 insert cartridge (Evans 

1960). Table 4-5 lists the intakes for an individual at EBR-I. An individual would have received these 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 22 of 39 

intakes only if present at the locations between November 2 and December 12, 1959. The intakes 

are for a total exposure period of 40 days.

4.2.2.3.4 Initial Engine Test 19(A) 

IET 19(A), conducted between February 9 and April 30, 1960, was a test series in the HTRE-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 the test were to: 

1. Operate the L2E-3 fuel cartridge at peak temperatures of 2,500° F and 2,600° F for 100 

hours or more at each temperature level to evaluate the effectiveness of the zirconiumdioxide coating against hydrolysis and the release of fission products. 

2. 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. 

3. Obtain additional information on the effectiveness of an electrostatic precipitator in 

removing fission products from the reactor effluent. (Pincock 1960)

Pincock (1960) summarized the estimated total fission product release for the test runs based on spot 

sampling and reported them as 10-minute-decayed curies. The total fission product release reported 

for IET 19(A) was 2,892 Ci. The release for this test was modeled as was that for DOE (1991). 

Table 4-5 lists intakes for EBR-I. An individual would have been exposed to these intakes only if 

present at the locations between February 17 and February 29, 1960. The intakes are for the total 

exposure period of 13 days. 

4.2.2.3.5 Initial Engine Test 25(A) 

IET 25(A), performed between November 15 and December 19, 1960, was an extension of the 

Phase II testing program in IET 18. The test was conducted in the HTRE-3 reactor configuration. 

Releases of airborne radioactivity that correspond 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)
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 23 of 39 

For this analysis the release was modeled, as in DOE (1991), as a straight-line trajectory such that 

the centerline plume affected EBR-I. Table 4-5 lists the intakes for this test. An individual would have 

received intakes only if present at the locations between November 22 and December 15, 1960. The 

tabulated intakes are for a total exposure period of 24 days. 

4.2.2.3.6 Initial Engine Test 26(A) 

IET 26 (A), conducted in HTRE-2, occurred between December 22, 1960, and March 31, 1961 (Field 

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 that exceeded 

120 kW. Releases for the IET 26(A) operation occurred from December 23 to 28, 1960. The insert 

under test was the L2E-6 cartridge, which consisted of fueled and nonfueled ceramic beryllium oxide 

hexagonal tubes coated on the inner surface with zirconium dioxide. 

The airborne release model was consistent with the model of DOE (1991) with an assumed straightline trajectory between TAN and EBR-I. Table 4-5 lists the intakes for EBR-I. 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 a total exposure period of 6 days. 

4.3 EXTERNAL DOSE 

External radiation dose at a facility can be created by direct radiation from two sources: direct beta/ 

gamma radiation from the facility or airborne effluents released from the facility or from adjacent 

facilities [21]. 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 [22]. A 

responsible H&S organization observes and curbs such trends to prevent personnel exposures from 

increasing unnecessarily. The following sections discuss facility fenceline film badge and TLD data 

that recorded doses from airborne fission product releases that had the potential for personnel 

exposure. Peterson (2004) contains more information on these two subjects. 

4.3.1 Facility Fenceline Annual Doses

Before 1970, many film badge or TLD measurements occurred 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 some of the highways at the southern end of the site. At first, the badges 

were retrieved and read once a month. The frequency changed to 6 weeks in 1962 and then back to 

monthly in 1963 (AEC 1963). Film badges were used through 9 months of 1966, and TLDs were 

used after that time. Beginning in 1967, TLDs were changed on a semiannual basis (AEC 1968). 

Recorded 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. Peterson (2004) contains more information and film badge data 

for this early period. The detection limit for the film badge reading was often quoted as 10 mrem for 

both beta and gamma readings (AEC 1963), and as 10 mrem for the TLD when it was first used. With 

the measured annual background radiation field at INL before operations began between 100 and 

150 mrem/yr, the monthly value of 8 to 13 mrem is at the detection limit of the film badge or TLD (i.e., 

within the deviation of the 10-mrem limit). Therefore, the uncertainty for monthly exchanges is higher 

than for less frequent exchanges. 

Facility and fence monitoring locations were established between the latter part of 1970 and the latter 

part of 1972. No film badge or TLD information is available for the EBR-I and BORAX locations when 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 24 of 39 

they were operational. From 1972 through 1983, facility fenceline TLD measurements, made on a 

6-month basis with five TLDs at each facility position, are available in the quarterly Environmental 

Monitoring Data reports (Walker 1972, 1973a,b, 1974a,b, 1975a,b,c; Dahl 1975; Sill 1976; Bills 1976; 

Williamson 1976a,b, 1977a,b,c, 1978a,b; DOE 1978a,b,c, 1979a,b,c,d, 1980a,b,c,d, 1981a,b,c,d, 

1982a,b,c,d, 1983a,b,c,d, 1984). Figure 4-5 (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 of 170 readings for a 6-month period. For this particular 2-year 

set of data, 1.5% of the 136 readings were assigned a 2-sigma uncertainty of 16% to 20%, and 18.4% 

of the readings were assigned a 2-sigma uncertainty of 11% to 15%. However, 80% of the 136 

values ascribed to the 34 locations over the 2-year period have a 2-sigma uncertainty of less than 

10%. 

Figure 4-5. Example of onsite TLD monitoring data (Williamson 

1976a, p. 38). 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 25 of 39 

To supply facility values for 1952 to 1972, the highest 6-month value from April 1972 to April 1973 for 

a facility was multiplied by 2 and applied to each year between 1952 and 1972. (Not all the listed 

facilities had commenced operation or even existed in 1952.) Facility fenceline TLD measurements 

could not be found for 1984 through 1992, but for 1993 and beyond such measurements were 

included in the EMRs (Mitchell 1994; Mitchell, Peterson, and Hoff 1995; Mitchell et al. 1996, 1997; 

Evans et al. 1998; Saffel et al. 2000; Stoller 2002a,b,c, 2003). For the period from 1984 to 1992 when 

TLD measurements were missing, reasonable interpolations were used to provide the missing values 

[23]. In addition, background TLD measurements that correspond with the facility fenceline TLD 

measuring periods were recorded in the EMRs. All reduced facility fenceline TLD data (facility 

fenceline data minus background) in the EMRs are listed in Table 4-6 (AEC 1971, 1972, 1973, 1974; 

ERDA 1975, 1976a, 1976b, 1977b; DOE 1978d, 1979e, 1980e, 1981e, 1982e, 1983e; 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, 2004, 2005). A more detailed discussion of the 

data is included in Peterson (2004). 

4.3.2 Facility Air Immersion Doses

ANL-W facility air immersion (beta-gamma) doses could be calculated from the noble gas and 

halogen portions of the operational releases and, if applicable, from the noble gas portion of the 

applicable episodic releases. This calculation should be unnecessary because these releases would 

be recorded in the fenceline TLD doses in Table 4-6. 

Table 4-6. ANL-W facility fenceline direct gamma values (TLD minus background) (mR) [25]. 

Year EBR-II TREAT Background 

1965–72 59 50 100–150 

1973 37 19 121 

1974 35 17 123 

1975 32 8 118 

1976 56 50 113 

1977 22 0 132 

1978 56 2 129 

1979 59 5 113 

1980 51 12 119 

1981 28 9 118 

1982 20 12 117 

1983 24 10 115 

1984 31 13 124 

1985 31 13 124 

1986 31 13 124 

1987 31 13 124 

1988 31 13 124 

Year EBR-II TREAT Background 

1989 31 13 124 

1990 19 13 124 

1991 19 13 124 

1992 19 13 124 

1993 28 16 111 

1994 15 3 130 

1995 17 7 116 

1996 22 21 129 

1997 16 16 128 

1998 0 11 131 

1999 13 13 122 

2000 25 26 129 

2001 0 3 140 

2002 18 39 120 

2003 20 11 117 

2004 23 10 118 

However, in considering this increased uncertainty, it is interesting to note the facility air-monitoring 

results that are also discussed in the annual EMRs. In each case, the facility air concentration is 

compared to the concentration 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. 

4.4 UNCERTAINTY 

The INELHDE (DOE 1991) contains a detailed discussion of the derivation of airborne releases for 

operational conditions and episodic events.
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 26 of 39 

4.4.1 Operational Releases

Discussions with the INELHDE (DOE 1991) authors suggest that operational releases, which were 

monitored, could be low by a factor of not more than 2. When the annual normalized ground-level 

concentration values are applied to the operational releases, the uncertainty could increase. 

4.4.2 Episodic Releases

As described in the INELHDE (DOE 1991), the episodic releases are a maximum reasonable value 

that is based on the amount of material available to be released and the conditions of the test. For 

such a release, 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 original effort to be “reasonably conservative” in the 

exposure estimates, some of the authors stated that the release that was considered for a particular 

episodic event might be low by as much as a factor of 3 [24]. 

4.4.3 Film Badge and Thermoluminescent Dosimeter 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 changeout (i.e., once per month, 

which was generally the case with film badges) [26]. The data for 1965 to 1972 in Table 4-6 is based 

on the highest 6-month TLD values of 1972 for the respective facility. Although the GE-ANP IETs 

were conducted in the late 1950s and early 1960s (the last IET, 26, 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 such that adjacent facilities were not affected [27]. 

However, after 1967 when facility fenceline measurements were routinely made with TLDs, with five 

TLDs at a given location the uncertainty is generally estimated at less than 10% [28]. Less than about 

20% of the time, these measurements have an estimated level of uncertainty as high as 20% [29]. 

4.5 ATTRIBUTIONS AND ANNOTATIONS 

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database. 

Norman Rohrig served as the initial Document Owner for this document. Mr. Rohrig was previously 

employed at INL which shared boundaries with ANL-W and used the same dosimetry systems 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. 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 27 of 39 

[1] Peterson, Henry. Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

Although major events at one facility have sometimes been measured at nearby facilities, they 

have never necessitated emergency measures. 

[2] Peterson, Henry. Principal Health Physicist. Intrepid Technology and Resources. 

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, 

which was not toward any other facility at the INL site. 

[3] Peterson, Henry. Principal Health Physicist. Intrepid Technology and Resources. 

September – December 2003. 

Because the ANL-W facility and personnel were moved from the southwest corner to the 

southeast corner of the INL site in 1966, it is good sense to determine the effluents that would 

be directed by meteorological diffusion to affect the facility in a new location. 

[4] Peterson, Henry. Principal Health Physicist. Intrepid Technology and Resources. 

September – December 2003. 

This is a recommendation to use the absorption types that are most predominant. This 

recommendation is supported by the reference Peterson 2004. 

[5] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

This recommendation to use a vapor intake of absorption type F material is based on the 

default value in ICRP Publication 66 (ICRP 1994). 

[6] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

This limited list of radionuclides provides more than 95% of the effective dose equivalent 

based on ICRP Publication 68 (ICRP 1995). Every radionuclide in the list of radionuclides 

from Peterson (2004) was multiplied by the effective dose coefficient for inhalation of 5-μm 

activity median aerodynamic diameter (AMAD) from Annex B of ICRP Publication 68 and then 

added together. In steps, nuclides were eliminated based on effective dose being less than 

0.1% of the total, then 0.3%, and then 0.4%. The fraction of effective dose equivalent lost is 

less than 5%, but the activity lost is considerably more. 

[7] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

An example of such analysis is Ebersole (1956). 

[8] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

The default geometric standard deviation of 3 has been confirmed by email, C. Bloom to N. 

Rohrig of 11/09/2006. 

[9] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

This statement is a conclusion from a review of the environmental data for onsite and distant 

offsite locations. 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 28 of 39 

[10] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. 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 trajectories 

was that only 16 releases had the potential to affect other INL facilities. 

[11] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

Through a review, conducted during the analysis described in DOE (1991), of the 

meteorological trajectories of all of the radiological releases, only nine of the releases affected 

the ANL-W facility – whether the ANL-W facility was to the southwest or at the southeast 

location, depending on the date of the release and location of the ANL-W facility. 

[12] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

The radiological impact analysis was performed to determine the diffusion at the EBR-I facility. 

[13] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

The radiological impact analysis was performed to determine the diffusion at the EBR-I facility. 

[14] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

“Fresh” fission products are those that result from a recent fission process; “aged” fission 

products are those that remain after a period of decay, which is measured in terms of months 

or years. A criticality accident that occurs in a solution of uranium fuel, such as those 

criticalities that occurred at the ICPP, involves aged/decayed fuel and, thus, aged or decayed 

fission products. However, the driving mechanisms for release are the heat and energy 

created by the criticality event, which create fresh fission products such as the short half-life, 

volatile noble gases and iodines that make up the majority of the fission product release from 

such an event. 

[17] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

The “episodic” releases ranged from criticality accidents that occurred in aqueous solutions of 

aged fuel to tests that involved burning fuel elements that had been allowed to decay for 

months, which had essentially no fresh fission products remaining in them. 

[16] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

The GE-ANP Program conducted tests in the HTRE cores that released fresh and aged fission 

products simply because the fuel cladding was destroyed and the fuel and the contained 

fission products were released directly to the atmosphere by way of the IET stack. 

[17] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

In the case of each episodic event, the release to the environment was a result of the actual 

operational history of the fuel involved being programmed into the RSAC computer program so 

that the actual or best-estimate fuel fission product inventory was used in the calculation of the 

release to the environment. In the analysis for DOE (1991), the actual or best-estimate 

release might have involved as many as 50 or 60 radionuclides, but the great majority of the 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 29 of 39 

resulting dose to a receptor was due to only a handful of radionuclides. The number of 

radionuclides was reduced as described in Peterson (2004) such that 95% of the original dose 

was preserved. 

[18] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

This limited list of radionuclides provides more than 95% of the effective dose equivalent 

based on ICRP Publication 68 (ICRP 1995). Every radionuclide in the list of radionuclides 

from Peterson (2004) was multiplied by the effective dose coefficient for inhalation of 5-μm 

AMAD from Annex B of 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. 

[19] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

This limited list of radionuclides provides more than 95% of the effective dose equivalent 

based on ICRP Publication 68 (ICRP 1995). Every radionuclide in the list of radionuclides 

from Peterson (2004) was multiplied by the effective dose coefficient for inhalation of 5-μm 

AMAD from Annex B of 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. 

[20] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

This limited list of radionuclides provides more than 95% of the effective dose equivalent 

based on ICRP Publication 68 (ICRP 1995). Every radionuclide in the list of radionuclides 

from Peterson (2004) was multiplied by the effective dose coefficient for inhalation of 5-μm 

AMAD from Annex B of 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. 

[21] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

The facilities at INL are spaced at distances such that direct radiation from one facility does 

not affect another facility. Therefore, the only other method for creation of a direct gamma or 

beta-gamma dose to personnel at the facility is for the facility to become engulfed in a cloud of 

RM, either from that facility or from another facility. 

[22] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

As a facility ages, dirt and waste, including RM, generally accumulates in the facility. A good 

example of such accumulation is at a foundry or machine shop. If management follows best 

practices, these materials and wastes are removed as they accumulate; if not, the general 

background radiation, as well as clutter in the facility increases with time. 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 30 of 39 

[23] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

This interpolation is described in Peterson (2004). For all facilities except TRA and ICPP, 

interpolated data is the average of TLD data for that facility from 1993 through 1999, zeros 

excluded. For TRA and ICPP, there is a downward trend in the readings with time. Therefore, 

for these two facilities, the bottom four values (1989 through 1992) are the averages of the 

seven values below, and the top five values (1984 through 1988) are the averages of the 

seven numbers directly above. 

[24] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

A factor of 3 was chosen by the authors of the DOE (1991) study as the upper bound value by 

which the GE-ANP releases could be inaccurate. 

[25] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

See the reference list. 

[26] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003.. 

Environmental film badges were exchanged monthly or more often in the early days. With 

normal background about 10 mrem/mo and a minimum reporting level of 10 mrem, only large 

excursions would be measurable. TLDs that were exchanged semiannually changed that 

situation appreciably. 

[27] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

There are multiple references to a test being authorized based on the prevailing and expected 

wind conditions. NOAA has been stationed in Idaho Falls to provide weather monitoring and 

test control. 

[28] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

This is based on Figure 4-5, which is a typical set of results. 

[29] Peterson, Henry, Principal Health Physicist. Intrepid Technology and Resources. September 

– December 2003. 

This is based on Figure 4-5, which is a typical set of results. 
Document No. ORAUT-TKBS-0026-4 Revision No. 01 Effective Date: 06/05/2007 Page 31 of 39 

REFERENCES 

AEC (U.S. Atomic Energy Commission), 1959, Annual Report of the Health and Safety Division - 

1959, IDO-12014, Idaho Operations Office, National Reactor Testing Station, Idaho, October. 

[SRDB Ref ID: 8080] 

AEC (U.S. Atomic Energy Commission), 1964, Health and Safety Division Annual Progress Report1963, IDO-12037, Idaho Operations Office, National Reactor Testing Station, Idaho. August.

[SRDB Ref ID: 13449] 

AEC (U.S. Atomic Energy Commission), 1960, Environmental Monitoring Data for the National 

Reactor Testing Station, Calendar Year 1959 and 1st Quarter of 1960, Idaho Operations 

Office, National Reactor Testing Station, Idaho. [SRDB Ref ID: 1286] 

AEC (U.S. Atomic Energy Commission), 1963, Environmental Monitoring Data (Report No. 12) for the 

First and Second Quarters, 1963, Idaho Operations Office, National Reactor Testing Station, 

Idaho. [SRDB Ref ID: 1301] 

AEC (U.S. Atomic Energy Commission), 1968, Environmental Monitoring Report No. 22 January-June 

1968, Idaho Operations Office, National Reactor Testing Station, Idaho. [SRDB Ref ID: 1353] 

AEC (U.S. Atomic Energy Commission), 1969a, Environmental Monitoring Report No. 23*, July - 

December and Annual Summary - 1968, Idaho Operations Office, National Reactor Testing 

Station, Idaho. [SRDB Ref ID: 1353, p. 7] 

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1361] 

AEC (U.S. Atomic Energy Commission), 1970a, Environmental Monitoring Report No. 25, July - 

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Station, Idaho Falls, Idaho. [SRDB Ref ID: 1361, p. 9] 

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June 1970, Idaho Operations Office, National Reactor Testing Station, Idaho. [SRDB Ref ID: 

1363] 

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Monitoring Program Report, January through December 1970, Idaho Operations Office, 

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Monitoring Program Report, January through December 1970, Idaho Operations Office, 

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National Reactor Testing Station, Idaho Falls, Idaho. [SRDB Ref ID: 1370] 
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AEC (U.S. Atomic Energy Commission), 1973, 1972 National Reactor Testing Station Environmental 

Monitoring Program Report, Idaho Operations Office, National Reactor Testing Station, Idaho 

Falls, Idaho, April. [SRDB Ref ID: 1459] 

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Monitoring Program Report, Idaho Operations Office, National Reactor Testing Station, Idaho 

Falls, Idaho, April. [SRDB Ref ID: 1480] 

Bills, C. W., 1976, "INEL Site Environmental Monitoring Data - First Quarter 1976," letter to J. A. 

Buckham, July 7. [SRDB Ref ID: 1517] 

Bloom, C., (ORAU), 2006, E-Mail Correspondence to N. Rohrig from C. Bloom: Environmental 

Internal Dose Distribution Assumption, November 9. [SRDB Ref ID: 30797] 

Dahl, A. H., 1975, "INEL Environmental Monitoring Data--Third Quarter 1975," letter to R. B. O'Brien 

(Aerojet Nuclear Company) and J. A. Buckham (Allied Chemical Corporation), U.S. Energy 

Research and Development Administration, Idaho Operations Office, Idaho National 

Engineering Laboratory, Idaho Falls, Idaho, December 23. [SRDB Ref ID: 1497] 

DOE (U.S. Department of Energy), 1978a, INEL Site Environmental Monitoring Data - First Quarter 

1978, Idaho Operations Office, Idaho Falls, Idaho. [SRDB Ref ID: 1539] 

DOE (U.S. Department of Energy), 1978b, INEL Site Environmental Monitoring Data - Second 

Quarter 1978, Idaho Operations Office, Idaho Falls, Idaho. [SRDB Ref ID: 1539, p. 35] 

DOE (U.S. Department of Energy), 1978c, INEL Site Environmental Monitoring Data – Third Quarter 

1978, Idaho Operations Office, Idaho Falls, Idaho. [SRDB Ref ID: 1539, p. 87] 

DOE (U.S. Department of Energy), 1978d, 1977 Environmental Monitoring Program Report for Idaho 

National Engineering Laboratory Site, IDO-12082 (77), Idaho Operations Office, Idaho Falls, 

Idaho, May. [SRDB Ref ID: 1543] 

DOE (U.S. Department of Energy), 1979a, INEL Site Environmental Monitoring Data – Fourth Quarter 

1978, Idaho Operations Office, Idaho Falls, Idaho. [SRDB Ref ID: 1539, p. 123] 

DOE (U.S. Department of Energy), 1979b, INEL Site Environmental Monitoring Data - First Quarter 

1979, Idaho Operations Office, Idaho Falls, Idaho. [SRDB Ref ID: 1548] 

DOE (U.S. Department of Energy), 1979c, INEL Site Environmental Monitoring Data - Second Quarter 

1979, Idaho Operations Office, Idaho Falls, Idaho. [SRDB Ref ID: 1551] 

DOE (U.S. Department of Energy), 1979d, INEL Site Environmental Monitoring Data - Third Quarter 

1979, Idaho Operations Office, Idaho Falls, Idaho. [SRDB Ref ID: 1557] 

DOE (U.S. Department of Energy), 1979e, 1978 Environmental Monitoring Program Report for the 

Idaho National Engineering Laboratory Site, IDO-12082(78), Idaho Operations Office, Idaho 

Falls, Idaho, May. [SRDB Ref ID: 1564] 

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