ORAU Team

NIOSH Dose Reconstruction Project

Technical Basis Document for the Oak Ridge National 

Laboratory -- Occupational Environmental Dose

Document Number:

ORAUT-TKBS-0012-4 

Effective Date: 05/07/2004

Revision No.: 00

Controlled Copy No.: ______

Page 1 of 41

Subject Expert: Robert E. Burns, Jr.

Document Owner

Approval: Signature on File Date: 05/04/2004 

Robert E. Burns, Jr., TBD Team Leader

Approval: Signature on File Date: 05/04/2004 Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 05/04/2004 Richard E. Toohey, Project Director

Approval: Signature on File Date: 05/07/2004 James W. Neton, OCAS Health Science Administrator

Supersedes:

None

TABLE OF CONTENTS

Section

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

Page

3

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

4.1 Introduction................................................................................................................................. 5

4.2 Historic Sources of Airborne Radioactive Material at ORNL........................................................ 5

4.3 Historic Onsite Environmental Monitoring Practices at ORNL ..................................................... 8

4.3.1 Airborne Radioactivity Measurements: November 1943 through March 1945 .................... 8

4.3.2 Concentration Measurements for Airborne Radioactive Material...................................... 10

4.3.3 Studies and Measurements of Onsite Particulates ........................................................... 13

4.3.4 Estimation of Radioactive Airborne Emissions ................................................................. 18

4.3.5 Onsite Exposure Rate Measurements.............................................................................. 20

4.4 Interpretation and Assessment of Available Monitoring Data .................................................... 21

4.4.1 Airborne Concentration Measurements............................................................................ 21

4.4.2 Radioactive Airborne Emissions....................................................................................... 25

4.4.3 Onsite Exposure Rate Measurements.............................................................................. 26

4.5 Derivation of Onsite Airborne Concentrations for Radionuclides of Concern............................. 27

4.5.1 Dispersion Coefficients for Principal Elevated Release Points ......................................... 28

4.5.2 Onsite Airborne Concentrations of 131I .............................................................................. 28

4.5.3 Onsite Airborne Concentrations of Mixed Fission Products.............................................. 31

4.5.4 Onsite Airborne Concentrations of Tritium........................................................................ 31
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4.6 Onsite Exposure Rate Data ...................................................................................................... 32

References ........................................................................................................................................35

4A Annual Airborne Releases Reported by Oak Ridge National Laboratory 

(curies)..........................................................................................................................

Attachment

39

4B Annual Average Airborne Concentration and Intake Data for Iodine-131, 

Tritium, and Particulate Mixed Fission Products at ORNL ............................................. 40

4C Annual Average Site-Wide Exposure Rates for Oak Ridge National 

Laboratory ..................................................................................................................... 41

LIST OF TABLES

Table

4-1 Principal elevated release points for airborne radioactive materials at ORNL...........................

Page

6

4-2 Dispersion coefficients used for ORNL airborne release points.............................................. 28

4-3 Bounding values of onsite airborne 131I concentrations at ORNL from RaLa 

processing .............................................................................................................................29

4-4 Bounding values of onsite airborne 131I concentrations at ORNL from short-decay 

thorex processing...................................................................................................................30

LIST OF FIGURES

Figure

4-1 Annual average gross beta concentration data from LAM and RAM networks for 

1956 through 1983.................................................................................................................

Page

24

4-2 Annual average exposure rates for ORNL.............................................................................. 34
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RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE

REV. 

NO. DESCRIPTION

Draft 12/29/2003 00-A New Technical Basis Document for the Oak Ridge 

National Laboratory Site – Occupational 

Environmental Dose. Initiated by Robert E. 

Burns, Jr.

04/14/2004 04/14/2004 00-B Incorporates internal review comments. Initiated 

by Robert E. Burns, Jr. 

05/07/2004 05/07/2004 00 First approved issue. Initiated by Robert E. 

Burns, Jr.
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 4 of 41

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AGL above ground level

Bq becquerel

CAM Continuous Air Monitor

cc cubic centimeter

CEDR Comprehensive Epidemiologic Data Resource

cfm cubic feet per minute

Ci curie

cm3 cubic centimeter

cpm counts per minute

CWS Chemical Warfare Service

d 24h-1 disintegrations per 24 hours

DOE U.S. Department of Energy

dpm disintegrations per minute

EEOICPA Energy Employee Occupational Illness Compensation Program Act

ft3 cubic foot 

GM Geiger-Mueller

HFIR High-Flux Isotope Reactor

hr hour

kW(t) kilowatts thermal 

LAM Local Air Monitor

LITR Low-Intensity Test Reactor 

m3 cubic meter

mA milliampere

MDA 

MDC 

MeV mega-electronvolt

MFPs mixed fission products

mph miles per hour

mR milliroentgen

mrad millirad

ORNL Oak Ridge National Laboratory

ORR Oak Ridge Reservation

PAM Perimeter Air Monitor

RaLa radioactive lanthanum 

RAM Remote Air Monitor
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4.1 INTRODUCTION

This technical basis document (TBD) describes the development of internal and external exposure 

data from onsite environmental sources of radioactive material for workers at the Oak Ridge National 

Laboratory (ORNL). Its purpose is to provide guidance to individuals performing dose reconstructions 

under the Energy Employees Occupational Illness Compensation Program Act (EEOICPA) for 

workers whose work history includes time at ORNL. The document addresses intakes of radioactive 

material that could have been present in the ambient air and ambient external exposures from 

process facilities, waste handling areas, and airborne noble gas emissions. 

The existence of site-wide monitoring data for external exposures dating back to early in the 

Laboratory’s history simplified the task of evaluating potential external exposures to unmonitored 

workers. However, despite the existence of substantial air monitoring networks both on the site and 

elsewhere, little information was available regarding airborne concentrations of the two principal 

nuclides of concern for inhalation exposure: 131I and tritium. Thus, a number of conservative 

approaches and assumptions had to be employed to develop the requisite concentration and intake 

data. These results should be considered more bounding than representative. 

Special Note Regarding Historical External Dosimetry Practices at ORNL

Dose reconstructors responsible for external dose assessments should note there are no known 

instances where control badges were stored or employed in such a way at ORNL where they would 

have been exposed to plant-related activity or otherwise utilized where excess exposure (i.e., 

exposure in excess of natural background) would have been subtracted from a worker’s dosimetry 

result. ORNL went to a “take home” badge (i.e., security badge and dosimeter combined) in the early 

1950s, and there are no indications in the records reviewed to date that the Laboratory ever employed 

control dosimeters for the purpose of subtracting a “background” from personnel exposures. It is 

known the Lab began subtracting backgrounds from take-home dosimetry results in 1989 or so, but 

the amount subtracted was based on controlled studies of offsite exposure rates. The signal 

subtracted was not based on control badge readings. Therefore, assigning external environmental 

dose (in addition to missed dose) should not be necessary for workers who were monitored with 

whole-body personnel dosimetry. The external environmental exposures addressed in this document 

are directed toward unmonitored individuals. 

The intake and exposure rate data given in Attachments 4B and 4C of this Technical Basis Document 

apply to the ORNL site as a whole. Technical Basis Documents and Site Profile Documents are 

general working documents that provide guidance concerning the preparation of dose reconstructions 

at particular sites or categories of sites. They will be revised in the event additional relevant 

information is obtained about the affected site(s). These documents may be used to assist NIOSH in 

the completion of the individual work required for each dose reconstruction. 

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

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

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

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

4.2 HISTORIC SOURCES OF AIRBORNE RADIOACTIVE MATERIAL AT ORNL

Table 4-1 lists the principal stacks that were in operation at ORNL over its history that were used to 

exhaust radioactive materials. The facilities and processes that contributed gaseous radioactive 

effluents over ORNL’s history changed somewhat with changes in the Laboratory’s mission and 
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facilities, coupled with improvements in air cleaning technologies. However, they can be placed in 

three general categories:

• Cooling air exhausted from the Graphite Reactor

• High-activity offgases from process vessels

• Low-activity ventilation air from hot cells and hoods

Table 4-1. Principal elevated release points for airborne radioactive materials at ORNL. 

Initial designation

Subsequent 

designation

Height in 

feet (AGL)

Years of 

operation Major processes served

105 3018 200 1943–1963 Graphite Reactor, LITR

205 (pilot plant stack) 3020 200 1943–present Dissolver offgases (pilot plant, RaLa facility), 

cell ventilation streams 

706-D (fan house stack) N/A 50 1944–1950 RaLa cell ventilation

900 (central offgas 

system stack)

3039 250 1950–present Isotopes Area, RaLa, vessel offgas, reactors, 

and hot cells. 

N/A 2026 75 1964–present High-Level Analytical Laboratory

N/A 7025 1967–ca. 1989 Tritium Target Fabrication Facility

N/A 7512 100 1965–1969 Molten Salt Reactor Experiment

N/A 7911 250 1966–present HFIR/TRU facility

The two original sources of airborne radioactive emissions at ORNL were the Graphite Reactor and 

the hot pilot plant (Building 205, later redesignated Building 3019). The gaseous wastes created by 

these two facilities were exhausted via the 105 stack and the 205 stack, respectively. The 105 stack 

ventilated cooling air exhausted from the Graphite Reactor, the principal radioactive constituent of 

which was 41Ar. The 205 stack exhausted offgases from dissolving and other chemical separations 

processes associated with the development of the bismuth phosphate process for separating 

plutonium from irradiated fuel. The principal radioactive constituents in this effluent stream were 

noble gases and 131I. 

The ORNL Graphite Reactor operated from November 1943 until November 1963. It was an aircooled graphite pile, with the exhaust cooling air ventilated via the 105 stack, which was later 

renumbered as the 3018 stack. The exhaust air system did not provide any holdup or delay, so the 

effluent stream included short-lived fission gases, their particulate decay products, and air activation 

products, most notably 41Ar. In addition, particulate mixed fission product activity would be released 

following incidents of ruptured fuel slugs. The Graphite Reactor was initially operated at a maximum 

power level of approximately 1,000 kW(t). However, it soon became apparent the system had been 

significantly over-designed, and modifications were made to allow its maximum power level to be 

uprated to approximately 5,000 kW(t). The estimated 41Ar emission rate from the 105 stack with the 

reactor at its initial power level was approximately 200 curies per day in 1944 (Parker 1944). The 

emission rate increased to approximately 500 curies per day in 1945 (Cheka 1945) following the 

uprate of the reactor’s operating power level. 

The exhaust air from the Graphite Reactor was discharged unfiltered until 1948 when a significant 

particle contamination problem was found on the Laboratory grounds. The Graphite Reactor was 

found to be one of the principal contributors to this particulate contamination, so a filter house was 

added between the exhaust plenum and the stack. The filtration system, which became operational in 

November 1948, consisted of parallel banks of roughing plus high-efficiency filters. Particulate 

releases from the 105 stack were largely mitigated once the filtration system was added. The 105 

stack also ventilated gaseous effluents from the Low-Intensity Test Reactor (LITR) from 1949 until 

1968. 
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Low volumes of high-activity offgases were produced from process vessels such as dissolvers, where 

the initial steps of chemical separations involving irradiated source materials were carried out. 

Examples of such operations include plutonium separations processes such as bismuth phosphate or 

PUREX and its variants, alternative fuel cycle processes such as Thorex, and production of 

radioisotopes such as 131I or 140Ba/La (i.e., the RaLa process). These high-activity offgas streams 

were ventilated via the 205 stack prior to completion of the ORNL central offgas handling system 

about 1950. After 1950, most of these high-activity offgas streams were ventilated via the 900 stack 

(later renumbered the 3039 stack) following treatment. Offgas treatment methods evolved and 

improved over the years and often included pretreatment (filters, scrubbers, etc.) at the generating 

facility prior to entering the central offgas system. 

High volumes of low-activity waste gases were produced by ORNL’s numerous hot cells and fume 

hoods. Most of these sources were ventilated locally prior to completion of the central offgas system 

about 1950. Following completion of the central offgas system, both filtered and unfiltered cell 

ventilation streams were exhausted via the 205 stack (later renumbered the 3020 stack). The filtered 

streams passed through a filter bank consisting of roughing plus high-efficiency filters prior to 

discharge. Unfiltered streams were discharged without additional treatment. 

As of the late 1940s, Cheka and McAlduff (1949) reported the principal sources of airborne 

radioactive emissions at ORNL to be:

• Cooling air exhausted from the Graphite Reactor via the 105 stack

• Emissions from the Redox process in the Pilot Plant (via the 205 stack)

• Dissolver offgas from 131I production in Building 706-C (via the 205 stack)

• Dissolver offgas from RaLa production in Building 706-D (via the 205 stack)

• Cell ventilation air from the RaLa process in Building 706-D (via the 706-D stack)

• Cell ventilation air from 131I and 135Xe production in Building 706-C (via roof vents)

Other sources of airborne radioactive material in this era were hot chemistry work conducted in hoods 

in Building 706-A. These were exhausted via roof vents. In addition, the tank farm was considered a 

source of airborne material from spills that subsequently dried and became airborne (Cheka and 

McAlduff 1949). In reality, studies in later years showed the W-9 tank in the south tank farm to be a 

significant source of local airborne contamination during waste transfer operations (Bradshaw and 

Cottrell 1954). This was presumably the case from the beginning, but was not recognized in the late 

1940s. Another source of airborne radioactive material cited for this era was fugitive emissions from 

the 706-D Building during RaLa operations (Cheka and McAlduff 1949). In fact, Cheka and McAlduff 

(1949) note that offgases from dissolver operations ventilated via the 205 stack seldom contributed to 

local air activity, whereas emissions from the latter parts of separations operations that were vented 

through the D-stack, roof vents, or the building itself were large contributors of activity. This situation 

was mitigated soon thereafter when the 900 stack and central offgas facility were placed in operation 

about 1950. 

Prior to 1950, airborne radioactive effluents from ORNL were discharged to stacks and vents serving 

individual facilities. However, the temporary nature of the Laboratory’s original mission resulted in it 

quickly outgrowing its waste handling measures once its status was changed to one of indefinite 

duration. Thus, in approximately 1950, ORNL installed and placed into operation a centralized offgas 

handling and treatment system as part of an effort to upgrade original, temporary laboratory facilities 

to those designed for long-term use. The central offgas system expanded over the years to 

accommodate expansion of the laboratory facilities and mission. A significant expansion in capacity 

of the system took place between 1956 and 1958. The primary stack used by the central offgas 

system was the 900 stack, which was renumbered as the 3039 stack around 1953. High-activity 
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gaseous effluent streams such as those from process vessels used in isotopic separations were 

exhausted via the 900 stack following treatment. High-volume, low-activity streams, such as those 

from hot cells and fume hoods, were exhausted via the 205 stack. 

Following completion of the centralized offgas facility in the early 1950s, two principal waste gas 

systems were in place at ORNL. The primary system handled high-concentration effluent streams 

such as those from chemical processing of materials such as irradiated nuclear fuels. This system 

discharged via the 250-foot 3039 stack after passing through various filters and precipitators. The 

secondary system was a common ventilation system for hot cells and laboratory hoods that 

discharged via either the 3020 or 3039 stack after passing through filter banks. The precise date that 

this system was completed is unknown, other than it was prior to 1959. The waste gas system for the 

Graphite Reactor remained unchanged and was independent of the two central systems. 

The vessel offgas treatment system of the central offgas facility originally consisted of a Cottrell 

electrostatic precipitator followed by high-efficiency filters. The precipitator consisted of 23 tubes, 

each 8 inches in diameter and 12 feet long. It operated at 52,000 volts and 130 mA, and was served 

by a continuous, recirculating-water flush system (Browder 1959). Additional treatment measures 

were added over time, such as scrubbers for radioiodines added in 1961 (Ohnesorge 1986). The 

report of the Applied Health Physics division for the first quarter of 1961 makes mention of the 

“installation of more efficient cleaning equipment for off-gas streams,” which appears to have occurred 

some time in 1960 (Hart 1961a). Charcoal filtration was added to the offgas line of the liquid waste 

storage tank used for waste from operations involving 131I about 1970 (Ohnesorge 1986). The Cottrell 

precipitator was eventually taken out of service, and demolition of the unit began in late 1981 (Auxier 

and Oakes 1982). The demolition of the precipitator was an element of a larger effort that began in 

October 1981 to upgrade 3039 stack services. Upgrade and removal efforts for the Laboratory’s cell 

ventilation system were also initiated during this era. 

4.3 HISTORIC ONSITE ENVIRONMENTAL MONITORING PRACTICES AT ORNL

4.3.1

Initial air monitoring efforts at ORNL focused on measurement of outdoor exposure rates from the 

radioiodine and noble gas emissions from the two stacks operating at that time: the 205 stack, which 

vented offgases from dissolving operations associated with the development of the bismuth 

phosphate process; and the 105 stack, which vented the discharged cooling air from the Graphite 

Reactor. Both of these stacks are 200 feet tall. The potential for internal exposure associated with 

radioiodine emissions was recognized, but it was felt that relatively little iodine was released from the 

Pilot Plant stack relative to the tolerance level of the day, which was 1 × 10-13 Ci cc-1 (Parker 1944). 

For comparison, this same reference states the tolerance levels (concentrations) for xenon and argon 

to be 2.3 × 10-11 Ci cc-1 and 1.8 × 10-12 Ci cc-1, respectively. Parker (1944) also states that selected 

staff members were checked for radioactivity of the thyroid gland with none being found, but neither 

the instrumentation used nor its sensitivity are discussed. The principal radioactive constituents of the 

effluent from the 205 stack were xenon and iodine. The activity in the 105 stack effluent was primarily 

41Ar. 

Airborne Radioactivity Measurements: November 1943 through March 1945 

Initial monitoring of outdoor airborne radioactivity at ORNL was performed using Lauritsen 

electroscopes to make exposure rate measurements on occasions when the plume from the 205 

stack reached ground level on the site. This monitoring method was employed from November 1943 

until March 1944 (Cheka 1945). 
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In May 1944, systematic monitoring of outdoor exposure rates on the ORNL site was begun using 

what was known as an X-22 ionization chamber (Cheka 1945). These units, sometimes referred to as 

“kangaroo chambers,” were condenser-type chambers based on the design of the pocket meters in 

use at the time. The chambers had 0.125-inch-thick bakelite walls and an active volume of 4,900 cm3

. 

They therefore had a relatively flat energy response and were essentially insensitive to beta activity. 

The chambers were calibrated using a radium gamma source (Parker 1944). A number of the 

chambers were modified to make them beta-sensitive by making 48 large holes in the walls and then 

covering the openings with graphite-backed cellophane or aluminum foil. The aluminum foil was 

preferred because it stood up better to weathering. Insulators for the collecting electrode in the X-22 

chambers were selected to limit leakage. 

In 1944, X-22 measurements were collected daily from chambers at eight locations. Some of these 

were on the site and some were in the surrounding area, out to about a mile away (Parker 1944). 

Cheka (1945) states X-22 chambers were placed at nine locations on and around the site by June 

1944. Of these nine locations, it appears that only the chambers located at the “rabbit cages” (800 

feet from the 205 stack) would have provided meaningful data about exposures to onsite personnel 

during that era. The location at White Oak Dam was not included in the daily monitoring routine. 

Each X-22 location had two stands designed to hold two chambers each (four chambers total) at a 

height of 5 feet off the ground. The intent was to have two of the standard gamma-only chambers and 

two of the modified beta-sensitive chambers at each location, with one of each pair acting as a control 

on the other. This was analogous to the Laboratory’s external dosimetry practice at the time, under 

which personnel were issued pairs of pocket meters, with one acting as a backup for the other in the 

event of a questionable reading. In practice, however, Cheka (1945) reported there were typically two 

gamma chambers deployed at each monitoring location, but the fact the modified (beta) chambers did 

not stand up well to weather made them unreliable. Thus, data from the beta-sensitive chambers is 

said to be “fragmentary.” 

Observation and trending of readings from the X-22 stations showed the principal source of gamma 

exposure on the ORNL site during this time was 41Ar discharged from the 105 stack and not 

xenon/iodine emissions from the 205 stack. [However, it was noted that when the bismuth phosphate 

process in the 205 building was shut down in early January, 1945, the monthly average exposure per 

X-22 station dropped 11% even though activity from the 105 stack increased by 8.5% (Wirth, Morgan, 

and Curtis 1945).] It was also noted that the radiation levels measured by the X-22 chambers were 

more dependent on their elevation than their distance from the stacks, with the exception of the 

station at the base of the stacks (which gave the highest reading of all). Diurnal variability was also 

noted, with the stations giving higher readings during the day than at night. The X-22 data also 

showed the ratio of the response of the beta-plus-gamma chambers to that of the gamma-only 

chambers to be independent of the radiation source (i.e., the chambers gave the same ratio whether 

they showed activity over background or not). Thus, site-related activity of that era does not appear to 

have had a strong beta component. 

In addition to monitoring via the X-22 chambers, film badges were deployed at locations further away 

from the site to provide monitoring in areas where the need for routine reading made use of the X-22 

chambers impractical. According to Parker (1944), the local X-22 chambers were read two or three 

times a day as of September, 1944. Film badges were placed at additional locations and were 

collected weekly starting on April 13, 1944. Parker states “regular film badges were used for this 

purpose and both beta and gamma ray measurements were obtained.” He goes on to say, “There 

has never been a measurable record in a one-week period,” which is not surprising given that the 

limit-of-detection for the film badges of that era was on the order of 30 mR. 
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Other methods of environmental monitoring at ORNL during its early period of operation included 

making exposure rate measurements in areas where it was evident, either visually or by odor, that the 

plume from the 205 stack had touched down locally (on the site). When this occurred, exposure rate 

measurements would be attempted using thin-walled Lauritsen electroscopes. Parker (1944) reports 

positive electroscope readings in three of eight such instances, with exposure rates of 1 to 2 mR hr-1

being observed for periods of 1 to 2 minutes. Grab sampling was performed early on using evacuated 

steel cans having a volume of approximately 10.5 liters. However, these samples could not be 

quantitatively assayed due to the lack of gaseous calibration standards. Precipitator sampling was 

used to determine natural radon and thoron concentrations. Using an assumed collection efficiency of 

70%, Parker (1944) reports the measured range of radon concentration to be 1 × 10-17 Ci cc-1 to 5 ×

10-16 Ci cc-1. The range of thoron concentrations measured was 8 × 10-19 Ci cc-1 to 1 × 10-17 Ci cc-1. 

Cheka (1945) described experiments performed in November and December 1944 to determine 133Xe 

concentrations in outdoor air downwind from the 205 stack. An accumulator was used to collect and 

liquefy 4,500 ft3 of air over the course of an hour. The liquefied air was then evaporated through a 

xenon trap, and the collected sample subsequently counted using a mica-window Geiger-Mueller 

(GM) tube. Fifteen runs were made at distances ranging from 150 feet to 3 miles from the plant, but 

most were made between 0.1 and 0.4 mile. The concentrations determined ranged from zero to 2.8 ×

10-11 Ci ft-3, and averaged 5.06 × 10-12 Ci ft-3. The highest values were obtained at 0.2 and 0.4 mile 

from the 205 stack. Note: 2.8 × 10-11 Ci ft-3 = 9.9 × 10-10 μCi cc-1. 

An experimental monitor consisting of a glass-walled GM tube and a scaler/recorder unit was installed 

approximately 800 feet northwest of the 105 stack in the latter part of 1944 (Cheka 1945). This unit 

began giving “fairly good” results in early 1945, with a maximum exposure rate of 0.009 mR hr-1 net 

(0.024 mR hr-1 gross) being measured due to 41Ar emission from the 105 stack during a period of 

southeasterly winds (i.e., blowing toward the monitoring unit). 

Cheka (1945) states that the program of monitoring via the X-22 chambers concluded on March 31, 

1945, as a “fairly complete” picture of the distribution of gaseous wastes obtained and it had been 

established that onsite exposure rates were within an order of magnitude of natural background. 

4.3.2

Following the conclusion of the X-22 program in March 1945, subsequent monitoring efforts for 

airborne radioactivity at the Laboratory switched in focus from external exposures to potential internal 

exposures from respirable materials (such as 131I and particulate matter). The date when they first 

appeared is unclear, but at some point prior to mid-1947, a system of Continuous Air Monitors (CAMs) 

began to evolve on the ORNL site. The initial complement of monitors consisted of three units. In 

October 1947, the three monitors were in the vicinity of Buildings 706-A, 115-B, and 735-B (Burnett 

1947a). They were in these same locations a year later in October 1948 (Brown 1948). These three 

units were the extent of the local air monitoring network until mid-1949, when the system was 

expanded to four monitors. In December 1948, the unit originally located at Building 115-B was 

moved to south of Building 105, and the unit originally located at 735-B was moved to west of the 

south tank farm (Cottrell 1948a). The system was subsequently expanded to 10 monitors around late 

1949. Note: A number of CAMs were employed for monitoring airborne radioactivity concentrations 

inside various ORNL facilities in addition to the three units originally installed to monitor outdoor air on 

the site. 

Concentration Measurements for Airborne Radioactive Material 

The CAMs were housed in louvered wooden enclosures, and drew air through a cylinder of filter 

paper inside a cardboard tube. The filter cartridge was shielded by a lead pig, and a GM tube was 

“fixed at its axis” (Cheka and McAlduff 1949). The GM detector was connected to a recording device 
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 11 of 41

(tape counters initially, then chart recorders as technology improved). The sample flow rate for the 

CAMs was reported by Morgan (1948) to be 7,000 ft3 per 24-hour day, which equates to roughly 5 

cfm. Cheka and McAlduff (1949) also assert a 5-cfm flow rate for these units. However, a memo 

describing an investigation of elevated airborne concentrations noted in July 1947 implies a flow rate 

of 3 cfm for that time (Burnett 1947b). In 1953, the filter media used in the CAMs is reported by 

Bradshaw and Cottrell (1954) to have been “H and V Number 70 paper,” where H and V stands for 

the Hollingsworth and Vost Company. The particulate filters used in the CAMs were eventually 

changed to glass fiber filters. 

Sometime around 1949, the ORNL Applied Health Physics Division began reporting data from the 10 

onsite CAMs in its periodic summary reports. CAM readings had been included in monitoring reports 

prior to this time, but only when instances of high air activity were noted, and even then the 

information was often qualitative. The data provided in the Applied Health Physics reports were

weekly average concentration data for the 10 onsite CAMs averaged over the reporting period (either 

quarterly or semiannually). The last report for a given calendar year gave the annual average of the 

weekly CAM results. It is not clear if the onset of reporting weekly average results for the CAMs in the 

Health Physics Division reports coincided with a change to a weekly exchange rate for the CAM 

filters. However, it is known the exchange rate was reduced to weekly at some point, and this 

appears to have happened prior to May 1951. In addition, the original scaler recording devices for the 

CAMs were changed to rate meters sometime in the early 1950s. Prior to this change, CAM results 

(i.e., in situ airborne concentration values) were derived from the number of counts integrated over an 

arbitrary time interval. The original scalar design, and the associated microswitches in particular, 

were a source of significant reliability problems for the original CAMs. The recording instruments 

were upgraded in December 1947 in an effort to improve reliability (Burnett 1948). The waste 

monitoring report for the week ending November 7, 1948, says the Instrument Division gave a delivery 

date of January 15, 1949, for the first of 10 logarithmic count rate meters to be used with the CAMs 

(Cottrell 1948b). However, the scalars were still in use through March 1949, but the transition to 

logarithmic count rate meters was completed by the early 1950s. 

During the first 6 months of 1956, an additional set of seven CAMs was activated. Six of these were 

placed at the U.S. Atomic Energy commission (AEC) Perimeter Guard Portals (gates). The seventh 

unit was on the Clinch River at Hickory Creek Bend. These new units included gummed-paper fallout 

trays and rainout collectors for measuring airborne particulates (fallout). The additional seven CAMs 

installed at the perimeter locations became known collectively as the Perimeter Air Monitors (PAMs), 

while the 10 monitors on the ORNL site became known as the Local Air Monitors (LAMs). The LAMs 

were outfitted with fallout trays and rainout collectors by the latter half of 1956. 

Along with the PAMs, another three CAM stations were installed sometime in the first half of 1956 at 

three remote locations: Berea, Kentucky; Corryton, Tennessee; and Kingston, Tennessee. These 

three stations were referred to as the Remote Area Monitors (RAMs). Data from the PAMs and RAMs 

were first reported in the Applied Health Physics semiannual report for the second half of 1956 (Hart 

1957). However, the RAM stations at Corryton and Kingston, Tennessee, were used only in 1956. 

Data are reported for the station at Berea, Kentucky, only in 1957 and 1958. 

Subsequently, an additional seven RAMs were activated at various Tennessee Valley Authority (TVA)

dams prior to 1959, bringing the RAM network to a total of 8 monitors for the calendar 1958 reporting 

period. These eight stations ranged in distance from the ORNL site from 12 to 120 miles. The RAM 

station at Berea, Kentucky, was retired by 1962, reducing the RAM network to seven stations ranging 

in distance from 12 to 75 miles from ORNL. A new RAM station at the TVA Power Service Center in 

northeast Knoxville, Tennessee, was added to the network in 1965 to bring the total back to eight. 
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Note: unlike the LAMs, which included GM tubes coupled first to scalers then to logarithmic count rate 

meters, PAMS and RAMs contained no monitoring equipment. 

The ORNL LAM network was upgraded beginning in 1962 by replacing the fixed filter with a step-type, 

moving-tape media (Morgan, Davis, and Hart 1963). The filter head housing the moving tape was 

monitored with an unspecified “beta-gamma monitoring device.” The impetus for changing the 

network from a fixed-filter medium to a moving tape appears to have been to provide greater 

versatility in the event of a radiological emergency. 

Starting in the second quarter of 1963, 12 new CAM stations began to be integrated into the LAM 

network. These monitors were added to expand coverage in response to a significant increase in the 

physical size of ORNL due to new facility construction (Davis and Hart 1963). However, data from 

only two of these stations were ever included in the network averages reported by the Health Physics 

Division. These two monitors were east of the 4500 building and at the HFIR complex. A 23rd LAM 

station, at Walker Branch, was added to the network by 1976. The results from this station were 

included in the network averages beginning with that year, bringing the total number of LAMs in the 

average to 13: the original 10 units plus the units east of the 4500 Building, at the HFIR complex, and 

at Walker Branch. The Walker Brach unit was approximately 2.5 miles east of the main Laboratory 

complex at the Low-Level Counting Laboratory. 

The PAM station at Hickory Creek Bend on the north side of the Clinch River was relocated in 1962 to 

the south side of the river and farther away due to the impending flooding of the Melton Hill Reservoir, 

scheduled for 1963 (Morgan, Davis, and Hart 1963). Two new PAM stations were added to the 

monitoring network in 1964, bringing the total number to nine. Another two stations were added by 

1983, making a total of 11. 

Radioiodine sampling capability was added to the PAM network in the second quarter of 1963 by the 

installation of activated charcoal cartridges in the air sampling line between the particulate filter and 

the air pump. The cartridges were removed weekly and assayed for radioiodine activity via gamma 

spectrometry (Davis and Hart 1963). Radioiodine (131I) concentration data measured via the PAM 

network were first reported for the fourth quarter of 1963. Radioiodine sampling capability was added 

to the LAM network by 1966. According to the Health Physics Division’s annual report for 1966, the 

MDA for the 131I sampling protocol employed in the LAMs and PAMs was 20 dpm, which 

corresponded to concentrations of 1.0 × 10-14 μCi cc-1 for the PAM stations and 2.0 × 10-14 μCi cc-1 for 

the LAM stations (Morgan, Snyder, and Davis 1967). ORNL practice at the time was to assign a 

concentration equivalent to 10 dpm for any charcoal cartridges showing a total amount of 131I less 

than the 20 dpm MDA (Morgan, Snyder, and Davis 1967). 

Beginning in 1975, the ORNL Health Physics Division began reporting isotopic analyses for filters 

collected from the LAM, PAM, and RAM stations. The results are reported as quarterly composites, 

but the specifics of how the analyses were performed or if all of the filters from all of the monitors were 

included are uncertain. The isotopic analyses show the dominant nuclide collected on the filters for all 

three networks to be 7

Be, which is formed naturally in the upper atmosphere by cosmic ray 

interactions with nitrogen and oxygen. Other nuclides observed on the filters, although in much lower 

concentrations, included 95Zr/Nb, 106Ru, 125Sb, 137Cs, and 144Ce (Auxier, Davis, and Turner 1976). No 

appreciable differences are seen in the isotopic mix data for the three air monitoring networks. By the 

early 1980s, as the airborne beta/gamma concentrations measured by the three air monitoring 

networks continued to drop, the relative concentration of 7

Be continued to climb to nearly 100% of the 

total. For instance, in 1979 7

Be accounted for 95.5% of the composite concentration for the LAM 

network, with fission product nuclides accounting for 4.1%. The difference between the two fractions 

is made up by small concentrations of primordial species. Fission product nuclides present in the 
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LAM samples for 1979 were 90Sr, 106Ru, 125Sb, 137Cs, and 144Ce (Auxier and Davis 1980). However, 

the 7

Be concentrations reported for calendar 1981 were significantly lower than those for previous 

years, implying a possible change in the analysis procedure. By 1977, filters collected from the LAM, 

PAM, and RAM stations were being analyzed for alpha radioactivity in addition to beta/gamma. Silica 

gel cartridges for tritium monitoring were added to selected LAM stations by 1980; however, no results 

(i.e., tritium concentration data) from this effort have been found in the references reviewed to date. 

By 1983, the filters were counted weekly for gross beta and gross alpha, and all filters for a given 

network (LAM, PAM, or RAM) were composited on a quarterly basis for specific nuclide identification. 

In response to changes in the regulatory framework placing increased emphasis on quantification of 

releases of pollutants from all pathways, a major reorganization of the various air monitoring networks 

in operation throughout the Oak Ridge Reservation (ORR) took place during 1985. As a result, the 

ORNL LAM network that had been in place for decades was retired, with the exception of a few 

stations that remained in service to form two new networks. One of these two new networks was 

established such that it covered the perimeter of the ORNL site. This new ORNL perimeter network 

consisted of four of the original LAM stations. The second new network was defined such that it 

covered the ORR perimeter. This new ORR perimeter network consisted of 11 stations: nine of the 

original PAM stations plus two of the original LAMs (MMES 1986). The 1985 reorganization does not 

appear to have affected the RAM network, other than all the air monitoring stations making up all the 

networks were renumbered during this time. However, annual average concentration data from the 

ORNL LAM network do not exist after 1984. 

4.3.3

With the recognition of the significance of particulate emissions at the Hanford Site in 1948, studies 

were conducted at ORNL beginning in May of that year to determine if a similar situation existed 

there. Indeed, a significant “particle problem” (as it became known) was confirmed, and this launched 

a program of monitoring for particulate radioactivity that continued for decades. Thus, there are 

voluminous data on particulate monitoring in ORNL’s onsite air monitoring history. However, the 

focus of the particulate monitoring program was quantification in terms of numbers of particles and 

fallout rates and, as such, the data are of limited use for estimating potential exposures to workers. 

For example, data from sedimentation frames were initially quantified in terms of particles per day 

(and later, particles per unit area), and air concentration data were expressed as particles per unit 

volume. In addition, the methods used to assess particle fallout rates and concentrations were 

preferentially biased toward larger nonrespirable particles. There is some information on particle 

activity distributions in the periodic monitoring summaries, but this information says nothing of particle 

sizes and thus is not useful for estimating inhalation intakes. This does not mean, however, that 

inhalation intakes of respirable particulates represent an unmonitored pathway. The numerous LAM 

stations positioned on the Laboratory grounds were designed to collect respirable particulate matter, 

and particulate air concentration data from these units are available as far back as 1947. 

Studies and Measurements of Onsite Particulates

This section of the TBD provides a summary of ORNL particulate monitoring practices, starting with 

the original recognition of the “particle problem” back in 1948. The data collected from the particulate 

monitoring program are of limited use for estimating inhalation intakes. However, the investigations 

into particulate emissions at ORNL are still of interest because they provide insight into issues such 

as sources of airborne particulates, different radionuclides that were released, remedial actions taken 

to reduce emissions, etc. This information was used as an aide to interpreting particulate data 

collected by the LAM network. In addition, the information was helpful in that it showed consistency 

between results from the onsite environmental monitoring program (for airborne concentrations and 

exposure rates) in the early years of Laboratory operation, and events that affected particulate 

emissions, such as installation of air cleaning measures or process upsets. 
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Following the recognition of the “particle problem” in May 1948, the Laboratory began an intense, 

large-scale program of investigations into the sources of particulate emissions, numbers of particles 

being released, and their physical nature. The initial phase of this effort began in June 1948, and by 

July of that year it had been shown that large (nonrespirable) particles could be found almost 

anywhere on the site, and that the principal source of these was the Graphite Reactor (Cheka and 

McAlduff 1949). This prompted Laboratory management to institute a large-scale program of 

measures to eliminate or control these particulate emissions in August, 1948 (Cheka and McAlduff 

1949). Particle control measures included installing filtration on the gaseous effluent streams 

subsequently found to be sources of particulate emissions, and dust (resuspension) control 

measures. 

Initially, the most problematic source of particulate emissions at ORNL was the Graphite Reactor. 

Specifically, the problem was high-specific-activity UO2 particles from ruptured fuel slugs being 

entrained in the cooling air exhaust stream. In June 1948, particles collected in the exit air duct of the 

Graphite Reactor (and thus upstream of the stack) over a period of 4 days showed particles ranging in 

size from 10 to 100 microns (Unknown 1948). Subsequent surveys around the site showed the 

largest particle concentrations (in terms of particles per unit area) were in the vicinity of the Graphite 

Reactor exhaust stack. These particles are reported to have ranged in size from 90 to 100 microns 

(Unknown 1948). However, Cheka and McAlduff (1949) report sizes ranged from approximately 80 to 

400 microns. Newman (1948) reported particles found by Health Physics being on the order of “a few 

tenths of a millimeter in diameter” and having beta count rates of 50,000 to 100,000 cpm. Follow-up 

investigations on particles collected in this initial phase of the large-scale investigation showed them 

to be UO2, but with a high percentage (in terms of alpha activity) of plutonium. Cheka and McAlduff 

(1949) reported that plutonium accounted for a weighted average of 88.9% ± 4.2% of the alpha 

activity. In addition, the particles contained mixed fission products and showed variable decay rates, 

implying material of different ages was being entrained and released. 

Following the discovery of large numbers of particles in the vicinity of the 105 stack, surveys in freshly 

graded areas and collections performed using sedimentation frames showed some particulate 

emissions did remain airborne. The recognition that the possibility of an inhalation hazard existed 

prompted corrective actions, principally, the installation of a filtration system between the Graphite 

Reactor exhaust air plenum and the stack. A filter house was constructed and put into operation on 

November 15, 1948 (Unknown 1948). Better surveillance methods for detecting and preempting slug 

ruptures were also initiated. 

Along with the particulate emissions from the Graphite Reactor’s cooling air exhaust, the second-most 

problematic source of onsite particulate emissions at ORNL was the RaLa process and its associated 

liquid waste handling operations. Particulate sampling of air exhaust streams from Redox operations 

(in Building 205) and RaLa operations (in Building 706-D) in 1948 showed the RaLa process 

produced radioactive particles “at a rate greater by several orders of magnitude than any other known 

source” (Unknown 1948). In contrast, the Redox process was deemed “a very minor contributor of 

radioactive particles” (Unknown 1948). Identification of the RaLa process as a major contributor of 

particulate emissions prompted corrective actions similar to those taken for the Graphite Reactor (i.e., 

installation of filters in the offgas streams). 

The nature of the particulate emissions from the RaLa process (specifically, from the cell ventilation 

air) was different than those from the Graphite Reactor. The particles from the Graphite Reactor 

exhaust air flow were UO2 containing plutonium and mixed fission products, and thus were discrete in 

nature. In contrast, Cheka and McAlduff (1948a) described particulate emissions from the RaLa 

process as more uniform in activity and in some cases “diffuse” in nature, “as if due to a liquid.” 

Cheka and McAlduff (1949) goes on to say that it is unclear if what they observed on the 
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autoradiographs were truly particles, or liquid droplets. They also point out that emissions from the 

RaLa process contained very few, if any, large particles, and the “particles” observed were presumed 

to be minute and of high specific activity. The particulate activity collected in the vicinity of the 706-D 

Building (RaLa process) was short-lived in relation to that collected in the vicinity of the 105 stack 

(Graphite Reactor). 

The studies of sources and magnitudes of radioactive particulate emissions at ORNL begun in May 

1948 led to increased monitoring of gaseous effluent streams and a program of routine surveillance 

and monitoring for particulate contamination. The particulate contamination and monitoring program 

was carried out by the Health Physics department, and initially included grab sampling, collection of 

particles using sedimentation frames, and the local CAMs. Grab sampling was performed using 

Filtrons, which drew air through a particulate filter at a rate of approximately 5 cfm. Sedimentation 

frames were used to monitor the rate of fallout of heavier particles that were contributing to ground 

concentration, whereas the Filtron and CAM data were used to determine particle densities in air 

(Cheka and McAlduff 1949). For each of these collection methods, the number of particles collected 

(either per unit volume or per unit area) was determined via autoradiography using X-ray film 

[reported by Cheka and McAlduff (1949) to be 14- × 17-inch Eastman “Blue Brand” film]. Morgan 

(1948) describes the process as a 24-hour exposure with each visible spot on the film counted. 

Morgan also states that this process would show particles ranging in activity from less than 100 to 

several thousand dpm. Cheka and McAlduff (1949) report the detection threshold for the 24-hour 

autoradiography to be equivalent to 0.8 count per minute with a mica-window counter in a 10% 

geometry. Note: The autoradiography method used for counting particles could only resolve the 

larger, higher-activity particles. 

In the case of the sedimentation frames, the particles were originally collected in a fiberglass mat in a 

3- by 4-foot frame. Following collection, the material in the mat was shaken onto an area sized such 

that it could be covered by a sheet of X-ray film (Cheka and McAlduff 1948b). However, it was later 

found that the fiberglass mat was retaining too much of the collected material, so it was replaced with 

heavy brown wrapping paper (Cheka and McAlduff 1948a). The process was again modified; rather 

than being shaken onto an area for radiography, the collected material was vacuumed using a 

“Sanitizor,” which was a vacuum cleaner with a replaceable filter paper bag (Cheka and McAlduff 

1949). The vacuuming technique is said to have done a better job of collecting the material adhered 

to wrapping paper (Cheka and McAlduff 1949). Counting the material in the filter bag worsened the 

detection threshold for the 24-hour radiography to 3.5 cpm equivalent (for a mica-windowed counter in 

a 10% geometry). 

With regard to particle sizes, Newman (1948) says experiments showed the Filtron units had a low 

efficiency for large particles (due to horizontal opening and low inlet velocity), but that those units 

should have been “reasonably efficient” for particles smaller than 10 microns. Morgan (1948) gives a 

sample flow rate for the Filtrons of approximately 300 cfm. However, Cheka and McAlduff (1948b) 

report the sampling rate as 300 ft3 per hour, or 5 cfm. The latter value is consistent with the flow rate 

reported for CAMs in both Morgan (1948) and Cheka and McAlduff (1948a). Newman (1948) asserts 

the sedimentation frames would probably not have been effective for capturing particles in the 5- to 

10-micron range, but should have been satisfactory for larger particles. A caveat is that 

sedimentation frames would have shown fewer particles as distance from the source increased, 

because larger particles would have settled out. Comparisons between particulate concentration data 

from collections using Filtrons and those made using Sanitizors in the fourth quarter of 1948 showed 

significant differences between the two. The Filtrons showed much higher concentrations than the 

Sanitizors, suggesting particulate emissions from chemical separations activities included a 

submicron component that passed through the Sanitizor filter bag. Studies conducted to determine 

solubilities of particulate material collected in the vicinity of the Graphite Reactor and Building 706-D 
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(RaLa operations) showed material collected near the reactor to be almost exclusively insoluble in 

water, whereas most material collected in the vicinity of Building 706-D was water-soluble (Cheka and 

McAlduff 1949). 

Other methods used for air sampling during this initial intense period of investigation into the “particle 

problem” included precipitrons, large air filtering units borrowed from the U. S. Public Health Service, 

and airway “Sanitizors,” which are described as a type of vacuum cleaner with a removable filter bag 

(Cheka and McAlduff 1948c). The Public Health Service units drew 240 cfm of air through 10 square 

feet of CWS 6 filter paper, which is asserted to have a collection efficiency of 99.97% “for the most 

penetrating particle size” (Cheka and McAlduff 1949). For the Sanitizor units, a flow rate of 56 cfm 

was used for calculation (Cheka and McAlduff 1949). The same reference states that the filter bag 

was estimated to have a collection efficiency of better than 95% for particles larger than 1 micron, and 

the inlet flow and collection nozzle were such it was unlikely that particles larger than 10 microns 

would be collected. 

Another element of the large-scale investigations of the “particle problem” was the start of an 

investigative nose swabbing program on September 17, 1948 (Cheka and McAlduff 1948a). 

However, the program was discontinued shortly thereafter (December 1948), “since no positive 

evidence of the presence of radioactivity in the lungs was obtained and no reliable quantitative 

measurements could be made” (Unknown 1948). As conceived, the nose swabbing program was to 

involve the selection of 50 individuals who would be requested to stop by the dispensary each day on 

completion of their shifts. Participation was good at first, but waned considerably over time, and this 

might have played a part in the decision to discontinue it. Of the individuals who did show up at the 

end of a given day, a nasal smear was collected using a cotton swab. The swabs were sliced open, 

counted in a “standard beta-gamma counter at 10% geometry,” and autoradiographed for 24 hours 

(Cheka and McAlduff 1948c). A total of 767 nasal smears was collected between September 17 and 

November 17, 1948, with particles detected on 66 of these (8.6%) (Cheka and McAlduff 1948c). 

Cheka and McAlduff (1948c) noted the percentage of positive nasal smears (for particles) correlated 

with the number of persons participating in the program for a given period. However, they also noted 

the incidence rate for positive nasal smears was 12.3% for persons working in the 105 Building 

(Graphite Reactor) versus a rate of 7.8% for other workers. The apparent incidence rate of 12.3% for 

105 Building workers came from a sample size of 138 smears, and the 7.8% rate from the remaining 

629 smears. The authors note the classic problem associated with nasal smear data, the 

interpretation of negative results. 

Following the initial, large-scale investigation into airborne particulate emissions at ORNL in the latter 

half of 1948, studies in 1953 documented by Bradshaw and Cottrell (1954) showed a significant 

correlation between particulate contamination and the RaLa process. However, the principal source 

of this contamination was not the 900 area stack (to which all RaLa gaseous wastes were vented 

starting in 1950). Rather, the primary pathway through which the RaLa process contributed to local

air contamination was the W-9 underground storage tank in the south tank farm. The tank, which 

received liquid wastes from RaLa separations, was vented to the atmosphere. Beta concentration 

measurements in the immediate vicinity of the tank vent were as high as 2 × 10-5 μCi cc-1 averaged 

over a 24-hour period between November 5 and November 25, 1953 (Bradshaw and Cottrell 1954). 

This period bracketed RaLa Run No. 54. Samples collected in this area prior to the first wastes being 

jetted to the W9 tank showed half-lives of greater than 136 days, and two principal components with 

beta energies of 0.35 and 3 MeV (Bradshaw and Cottrell 1954). In contrast, samples collected during 

various stages of the RaLa run showed initial half-lives of 8.5 to 22 days. The same samples showed 

half-lives of 21.5 to 58 days after approximately 1 month of decay. Radiochemical analyses were 

performed on some of these samples and are reported in Bradshaw and Cottrell (1954, Table VI). 

These analyses showed roughly 50% to 65% of the activity to be from total rare earths (cerium et al.), 
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with the next largest constituent (on an activity basis) being strontium, primarily 89Sr. Other 

constituents included ruthenium, barium, zirconium, niobium, and tellurium/iodine. The difference in 

total activity collected on filter paper media versus that collected using a bubbler sampler indicated 

“that the major portion of the activity escaping the W-9 tank was in particulate form” (Bradshaw and 

Cottrell 1954). 

Bradshaw and Cottrell’s data prompted a subsequent experiment during RaLa Run 57 (in July 1954) 

to assess the effectiveness of a fiberglass filter on the W-9 tank vent (Bradshaw and Cottrell 1954). 

The filter reduced local air concentrations in the vicinity of the tank vent by about 2 orders of 

magnitude over those measured during RaLa Runs 52 and 54. However, a bigger reduction in local 

air activity was achieved by sealing off the opening around the tank liquid level indicator. Other 

openings in the tank were sealed prior to the start of the experiment. Note: It is not clear from the 

report if the seals on the tank openings and the filter on the vent line were left in place after the 

experiment was completed. 

A significant increase in particle concentrations detected by the onsite CAMs in the first half of 1956 

prompted an investigation into the cause. It was found the filters downstream of the Cottrell 

precipitator in the central offgas system serving the 3039 stack had been eaten away by acid fumes 

(Hart 1957). The filters were replaced and a “considerable decrease” in particle concentrations was 

subsequently noted (Hart 1957). 

By June 1953, the methods used to measure particulate fallout rates and concentrations on the ORNL 

site consisted of the network of 10 LAMs, sedimentation trays (then referred to as collecting frames), 

and four U. S. Public Health Service air samplers. However, by July 1953, it appears that use of 

collecting frames had been discontinued, and only particulate concentration data were reported in 

terms of particles per 1,000 ft3 of air. Data from the collecting frames were reported on a rate basis, 

as particles per 24 hours. Starting in the fourth quarter of 1954, particulate data from the 10 LAMs 

were reported in terms of concentration in particles per 1,000 ft3 of air and in terms of an activity 

distribution. Activity was determined by comparing autoradiograph images from field samples to 

those from a particle of known activity. Activity was reported in terms of disintegrations per 24 hours. 

The activity distributions were given as numbers of particles in each of four activity ranges: less than 

105 d 24h-1, 105 to 106 d 24h-1, 106 to 107 d 24h-1, and greater than 107 d 24h-1. (These distributions 

do not provide information on particle sizes beyond what is known about the physical characteristics 

of the LAMs and the filter papers employed.) By the first half of 1955, use of the Public Health 

Service samplers appears to have been discontinued, leaving the onsite LAMs as the sole means for 

determining onsite particulate concentrations. This was the case until sometime in the first half of 

1956, when gummed-paper fallout trays were collocated with new PAM stations at the AEC perimeter 

guard stations and at three remote locations. The onsite LAMs had fallout trays added, and collectors 

for measuring fallout in rainwater were added to all of the CAM stations during this time. 

The gummed-paper fallout trays deployed at the air monitoring stations are described as cellophane 

sheets 1 square foot in area with a gummed surface on one side. The sheets were mounted on an 

aluminum tray with the gummed side facing up (Hart 1957). The sheets from the fallout trays (LAMs, 

RAMs, and PAMs) were collected weekly and assayed in terms of microcuries per square foot. 

The material collected in the gummed-paper fallout trays was assayed first by autoradiography and 

then by beta/gamma counting. Autoradiography was performed by covering the gummed paper with 

a sheet of rubber hydrochloride, cutting it from the collection frame, and mounting it on a sheet of 

Kodak Blue X-ray film (Browder 1959). The film was exposed in a dark room for 24 hours, after which 

spots created by radioactive particles were counted via the unaided eye. Activity ranges for the 
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particles were determined by comparing the diameter of the spots to those created using particles of 

known activities. 

Following radiography, the material collected on the gummed paper was beta/gamma counted by 

folding the paper into a small square, placing it in a crucible, and ashing it in a muffle furnace. The 

ash was transferred to a stainless-steel counting dish and counted in a gas-flow proportional counter 

(Browder 1959). Activity was asserted in microcuries and results were reported as microcuries per 

square foot. 

Experiments performed in 1953 using Millipore filters suggested the autoradiographic methods and 

filter papers used in the LAMs to assert average airborne particle concentrations suffered from low 

bias due to particles embedded in the LAM filters and the method used to count spots on the exposed 

X-ray films (Bradshaw and Cottrell 1954). 

The practice of radiographing the material collected from the gummed-paper trays was eventually 

discontinued, and the fallout data from the trays was reported only in terms of activity per unit area. 

The gummed-paper fallout trays and the rainout collectors remained part of the complement of 

collection media utilized by the ORNL onsite air monitoring network through 1984. 

4.3.4

The initial emphasis on effluent monitoring at ORNL was on liquid waste discharges, either directly or 

indirectly, to White Oak Creek. Early data on gaseous effluent discharges are scarce, as airborne 

releases were not considered a significant contributor to on- or offsite contamination or personnel 

exposures. The combination of airborne concentration data from the CAM stations and the numerous 

exposure rate measurements indicated minimal impact from airborne releases. Therefore, the focus 

of stack monitoring practices for the first decades of Laboratory operations was the identification of 

off-normal events resulting in abnormally high stack effluent concentrations. The original stack 

sampling methods focused on sampling offgas streams more so than monitoring, with samples drawn 

through filter media for subsequent counting for gross beta and gross alpha. Counting results were 

divided by the volume sampled to yield average concentration over the sampling interval. Gaseous 

releases were quantified in this manner until the early 1960s, when the Laboratory began reporting 

131I releases in total curies. However, quantification of other components of the Laboratory source 

term did not begin until the early 1970s, when the Laboratory began reporting discharges for noble 

gases (in 1969), tritium (in 1972), and “unidentified alpha” (in 1972). Improvements to stack 

monitoring practices in 1970 allowed the Libratory to report separate release estimates for 85Kr and 

133Xe rather than for total noble gases. 

Estimation of Radioactive Airborne Emissions

From an offsite dose perspective, the most significant airborne radioactive effluent from ORNL was 

131I from the period 1943 until around 1972. The importance of 131I to collective offsite dose 

diminished after 1972 due to a decrease in production and improvements in the offgas treatment 

systems. As a result, tritium became roughly equal to 131I in terms of contribution to collective dose in 

the period from 1972 to 1979 (Ohnesorge 1986). After 1979, 131I fell out of the picture, leaving tritium 

as the principal contributor until 1997, when tritium releases dropped to a point where they were no 

longer significant. At this time, 41Ar releases from the HFIR (7911 stack) became the principal 

contributor to offsite collective dose, supplanted with 138Cs if the HFIR was not operating for extended 

periods. Note: these materials are discussed in terms of contribution to offsite dose since that was 

the driver for the effluent monitoring and reporting program. This is not meant to imply that different 

materials were important contributors to onsite dose to ORNL workers. 
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By the late 1950s, approximately 90% of the gaseous radioactivity generated by the Laboratory was 

discharged via the principal stack, referred to as 3039 (also referred to in older literature as the 900 

area stack) (Manneschmidt undated). The other contributors to the site’s airborne source term in this 

era were the 3018 and 3020 stacks, formerly known as the 105 and 205 stacks, respectively. The 

3039 stack discharged gaseous wastes from the isotope processing areas, the ORNL Research 

Reactor, and various buildings housing research and development groups. The 3018 (reactor area) 

stack vented cooling air discharged from the Graphite Reactor, consisting primarily (on an activity 

basis) of 41Ar. The 3020 (pilot plant) stack discharged ventilation air from hot cells and operating 

areas where processing of irradiated reactor fuel was performed. 

Routine monitoring of the gaseous waste discharges at the lab began in 1956 when the Health 

Physics Department installed a filter tape monitor on the Graphite Reactor (3018) stack 

(Manneschmidt undated). A similar device was installed on the pilot plant (3020) stack in 1957, and 

coverage was extended to the principal stack (3039) in 1959. Manneschmidt asserts nonroutine grab 

sampling was performed prior to the installation of these devices. The tape monitors did not provide a 

measurement of activity discharged, but simply a means to detect unusually high activity (via a sidewindow GM tube). Manneschmidt reports that tapes were changed daily, and used tapes were 

counted for gross beta/gamma and gross alpha activity following 72 hours of decay. No nuclide 

identification was performed. These systems suffered from significant sample line losses. 

A report prepared by ORNL for public hearings on radioactive waste disposal before the Joint 

Congressional Committee on Atomic Energy (January 28 through February 3, 1959) provides more 

detail on the stack monitors in use at that time. The system for the 115 (Graphite Reactor) stack is 

consistent with the description by Manneschmidt (Undated), and the collection media was HV 70 

paper (Browder 1959). The systems for the 205 and 900 stacks are described as sample lines 

drawing the extracted air through Millipore filters. Unlike the system for the 105 stack, no mention is 

made of count rate meters or chart recorders. Browder (1959) reports the filters were changed daily 

and counted for gross beta and gross alpha activity. No mention is made of how long the filters were 

allowed to decay prior to counting or if any sort of decay correction was performed. Results obtained 

were compared with MPC values for gross beta and gross alpha. 

Gaseous waste data reported for the second quarter of 1961 were in terms of “long-lived” gamma 

activity for the principal Laboratory stacks. These data do not include, for instance, the contribution 

from 41Ar discharged from the 3018 stack (Cox 1961). However, 131I is included, so it is not clear how 

long the filter papers were allowed to decay before counting (or if the assay results were corrected to 

account for decay during collection and between collection and counting). Cox (1961) asserts that 

90% of the total long-lived beta/gamma activity discharged came from the 3039 stack, and that 50% 

of this activity came from iodine processing operations in the isotopes area. Principal components of 

the activity are 131I, 133I, and traces of 103Ru. The same report states the use of paper filters had been 

discontinued in favor of membrane filters, with the latter giving better collection efficiency, and 

retention characteristics that made for higher alpha counting efficiencies. 

Reliable estimates of quantities of airborne radioactive effluents discharged by ORNL are not 

available prior to the installation of the sampling system at the 50-foot level of the 3039 stack in the 

Fall of 1961 (Jordan 1974). The monitoring equipment for the 3039 and the 7911 stacks was modified 

in April 1969 to provide quantitative information on noble gas releases for the first time. Noble gas 

releases were initially reported in terms of 133Xe equivalence, with this practice transitioning to 

reporting equivalent quantities of 133Xe and 85Kr shortly thereafter. Either way, ORNL felt the reported 

releases were quite conservative. 
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 20 of 41

Quantities of tritium released to the atmosphere asserted by ORNL are upper bound estimates based 

on differences between material received and the sum of material sold plus that held in inventory. 

This method was necessary due to the lack of reliable tritium monitoring equipment at the largest 

source of the Laboratory’s airborne tritium emissions, Building 3033. This facility received tritium 

primarily from the Savannah River Plant, purified it, and repackaged it for industrial and research 

applications. The tritium released in the Building 3033 gaseous waste stream was discharged via the 

3039 stack. The 3033 building did have monitoring equipment in its offgas system, but the monitor 

could not differentiate between tritium and the 85Kr that was also processed there. Another source of 

atmospheric tritium emissions at ORNL was the Tritium Target Fabrication Facility (TTFF), Building 

7025. This facility operated intermittently in relation to Building 3033 and handled tritium in much 

smaller quantities. Tritium was released from the TTFF via its own local stack. 

Summary data of annual discharges of radionuclides to the atmosphere began appearing in the 

annual reports of the Health Physics Division in 1976. Discharge estimates were provided for 

individual stacks for tritium, 85Kr, 131I, and 133Xe (Auxier and Davis 1977). 

By 1984, noble gas discharges via the 3039 stack were quantified via “a real-time (continuous) 

monitor with an electronic integrator” (MMES 1985). Prior to this time, discharges were estimated via 

manual integration of stripchart data. Tritium discharges, 99% of which were via the 3039 stack, were 

quantified for the first time using a real-time monitor for that release point. All tritium discharge 

estimates asserted prior to 1984 are based on inventories. Remaining tritium releases (from the local 

stack at Building 7025) were quantified via silica gel samplers. Releases of alpha and beta 

particulates and iodine were quantified using filters and charcoal cartridges, respectively. These 

media were collected three times a week from the 3039 stack sampling system and weekly for other 

stacks. Iodine-131 was released primarily from the 3039 and 7911 stacks from operations involving 

fuel elements and isotope production activities. Radon-220 was released from the 7911 stack from 

storage of a neutron-irradiated 226Rn source at the HFIR complex. The source was sealed in 1983, 

but radon releases continued from contamination of the storage area (MMES 1985). 

A summary of total annual airborne releases of 131I, tritium, 85Kr, and 133Xe reported by ORNL is 

included as Attachment 4A to this document. All values are given in total curies discharged to the 

atmosphere. The data date to 1961 for 131I releases, 1969 for tritium releases, and 1970 for noble 

gases. For completeness, data for 41Ar and 138Cs have been included for later years when these 

nuclides represented the maximum contributions to collective offsite dose. 

4.3.5

By the late 1940s, ORNL had an established program of routine measurements to monitor the 

ambient “background” at the Laboratory. Measurements were made at approximately 50 specific 

locations in the Laboratory area using a portable GM counter and scaler. The measurements, 

referred to as “background checks,” were performed approximately monthly and were originally 

quantified in terms of count rate. Starting in February 1949, a radium source was used to determine a 

calibration factor in cpm per mR hr-1 for the detector used, and results from then on were reported in 

mR hr-1. However, data for 1947 and 1948 have been found in which the results have been converted 

from cpm to mR hr-1 (Morgan 1959). 

Onsite Exposure Rate Measurements

The GM counter used for background measurements was mounted in a 2- by 2-inch piece of lumber 

that extended approximately 3 feet from the bed of a pick-up truck (Cottrell 1949). This arrangement 

placed the counter approximately 3 feet off the ground. The detector was coupled to a scaler and a 

portable power supply (if AC power wasn’t available at the measurement location). Counting intervals 
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 21 of 41

ranged from 2 to 5 minutes. Note: The exposure rates at some measurement locations were such 

that the GM tube could not be used. In these instances, an electroscope was used. 

It is suggested in some of the periodic reports of the ORNL Health Physics Division that a scintillation 

detector was sometimes used rather than a GM counter for the background measurements, but the 

protocol appears to have remained otherwise unchanged until 1962. Experimental, weather-proofed 

film packets were tested for continuous environmental monitoring during 1961 and found to be 

satisfactory. Morgan and Hart (1962) report the results from the film packets differed from survey 

meter results by less than 10%, and thus the film packets became the principal means for monitoring 

background exposure rates beginning with 1962. The packets were placed near the original stations 

where monthly background readings had been taken. The packets were processed quarterly. It 

appears that onsite ambient exposure rate measurements were performed using film packets through 

1970. Annual average exposure rate data do not appear in the Health Physics Division reports for 

1971 through 1975, and have not been found in other references reviewed to date. 

By 1973, the practice used by ORNL for measuring exposure rates at offsite locations in East 

Tennessee changed from static measurements using GM or scintillation detectors to the deployment 

of thermoluminescent dosimeters (TLDs). The TLDs were placed at the PAM and RAM stations at a 

height of 4 feet off the ground (Auxier, Davis, and Turner 1974). The thermoluminescent material was 

calcium fluoride (Harshaw TLD-200). Control badges were irradiated to 20 mR at the midpoint of a 

deployment period to provide a fade correction (Auxier, Davis, and Turner 1975). The TLD 

measurement protocol for the PAM and RAM stations was modified slightly in subsequent years, with 

the dosimeters suspended at a height of 1 meter off the ground and the exchange frequency for the 

dosimeters at the RAM locations extended to semiannually (Auxier, Davis, and Turner 1976). The 

exchange frequency for the dosimeters at the PAM stations remained monthly. 

TLDs were included at the onsite LAM stations as of 1976. Commensurately, annual average 

exposure rates on the ORNL site were again reported in the summary reports of the Health Physics 

Division. By 1982, the dosimeters used for environmental exposure measurements were changed to 

a combination of lithium fluoride and calcium fluoride units, suspended 1 meter off the ground. The 

exchange frequency for the dosimeters remained the same: monthly for PAM stations and 

semiannually for LAM and RAM stations (ORNL 1983). The practice of reporting onsite average 

exposure rates for ORNL appears to have ceased after 1984, presumably an artifact of the major 

reorganization of the air monitoring networks in operation throughout the ORR that took place during 

1985. 

4.4 INTERPRETATION AND ASSESSMENT OF AVAILABLE MONITORING DATA 

4.4.1

Until 1966, measurements of airborne radioactivity concentrations for local onsite air at ORNL were 

performed by sampling with particulate filters alone. Thus, while results from the Laboratory’s onsite 

air monitoring network should be reliable for mixed fission product activity, the results were not 

deemed reliable for estimating 131I concentrations for dose reconstructions under the EEOICPA. 

Iodine-131 was a significant constituent of ORNL airborne releases from early in site history until the 

early 1970s; however, charcoal cartridges were not added to the sampling train of the local air 

monitoring network until 1966. Similarly, Laboratory onsite air monitoring results in the periodic 

reports of the Health Physics Division do not address tritium concentrations. Tritium became a 

relatively significant constituent of airborne releases once 131I releases dropped off following a 

reduction in isotope sales. The largest airborne tritium releases from ORNL occurred between 1980 

and the early 1990s. 

Airborne Concentration Measurements
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It appears the protocol for determining the long-lived airborne beta concentration from the CAM filters 

(from all three monitoring networks) remained fairly consistent over the Laboratory’s history. As 

described in ORNL-2601 (Browder 1959), CAM filters were changed weekly unless significant 

contamination occurred. For the LAMs, which included GM detectors and count rate meters, count 

rate data were tabulated hourly, with air concentrations indicated by the rate of increase of the count 

rate (under the assumption that the half-life of the activity collected was long in relation to collection 

time). Following collection, the filters were analyzed first by autoradiography, then by gross beta 

counting. Autoradiography was done by placing 12 filters at a time on a 14- by 17-inch sheet of 

cardboard and placing the cardboard in direct contact with a sheet of Kodak Blue X-ray film (Browder 

1959). This filter/film arrangement was exposed for 24 hours in a dark room. Afterward, spots (i.e., 

particles) were counted via the unaided eye, with particle activities assigned by comparing the 

diameter of the spots to those created using particles of known activity. 

Following radiography, the CAM filters were counted for beta/gamma activity by placing them inside a

1-inch-diameter aluminum cylinder containing an axially mounted GM tube (Browder 1959). This 

geometry essentially replicated that of the LAMs in the field. The activity (in microcuries) was 

determined and divided by the volume sampled to get average concentration. The ORNL-2601 report 

(Browder 1959) mentions performing gamma spectrometry and chemical separations in instances 

where high activity was encountered to try to determine its source. However, it is not clear if this was 

done for the CAM filters or just material collected on the gummed-paper fallout trays deployed at the 

CAM stations in 1956. 

In the process described by Gupton (1973), air samples were collected using 6- by 6-inch HV-70 

filters at a flow rate of approximately 5 cfm for 7 days. After 7 days, the filters were collected and 

replaced. After approximately a week of decay, the filters were counted using a beta scintillation 

detector with an efficiency of 0.5 for 90Sr/Y. If the apparent concentration was low, the result based on 

the 90Sr/Y efficiency was retained, and all activity was considered “long-lived.” However, if the result 

was elevated, the isotopes were identified and an appropriate efficiency applied. The MDC of this 

technique for respirable activity based on the 90Sr/Y efficiency was “about 5 × 10-15 μCi/ml” (Gupton 

1973). Review of the periodic waste monitoring reports from the late 1940s suggests filters were 

collected on a shift basis (every 8 hours), and thus the MDC would increase commensurately. Cottrell 

(1948a) reports the background of a CAM after an hour of operation was about 300 to 800 cpm. How 

long the filters were allowed to decay prior to counting and the specifics of how they were counted 

during this era are not known. In addition, by the late 1940s, the filters were autoradiographed to 

determine the number of discernible particles collected. Alpha counting was not performed, though 

special sampling for alpha air activity could be conducted on an as-needed basis using grab 

samplers. 

Note: The gross beta results reported for the ORNL CAM stations have always been in terms of “longlived” activity, meaning the filters were decayed for long enough that short-lived particulate decay 

products of noble gases and radon were not included in the results. 

The average concentration reported for the LAM network for 1949 through 1951 is notably higher than 

for all subsequent years, with the exception of 1954, when the average is driven by the RaLa accident 

that occurred that April (see the summary of gross beta results from the LAM network in Attachment 

4B). Thus, the expansion of the LAM network from four units to ten that occurred around late 1949 

does not appear to have had a significant impact on the site-wide average. This implies the network 

was providing representative coverage. The reason for the nearly order of magnitude drop in reported 

concentrations between 1951 and 1952 is not known, but it might have been the result of a change in 

analytical methods or assumptions. The electronics used in the LAMs were switched during this time 

from scalers to ratemeters, but this is not believed to have affected reported average concentrations. 
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Referring to emissions from the 900 (3039) stack during RaLa runs 52 and 54, Bradshaw and Cottrell 

(1954) report “No correlation was found between the stack activity and the response of the CAMS on 

the Laboratory area in either run.” They go on to add that good correlation was observed between 

airborne emissions from the vent of the W-9 underground waste storage tank and the response of the 

CAMs. Thus, ground-level radioactive material concentrations in air on the ORNL site were primarily 

due to ground-level and low-elevation release points, and not stack emissions. That was especially 

true until about 1970, when charcoal filtration was added to the offgas line of the liquid waste storage 

tank used for waste from operations involving 131I. 

Though it is doubtful the LAMs ever provided an accurate representation of onsite ground-level 131I 

concentrations prior to the addition of charcoal cartridge sampling media in 1966, the filter paper 

media did collect some radioiodine activity. Presumably the filter media would collect iodine-bearing 

aerosols with particle sizes in the respirable range, and airborne radionuclide concentrations were 

treated as 131I equivalent (i.e., were compared to the MPC for 131I) in the early years of Laboratory 

operations. Decay curves produced using CAM filters in service during a period of increased air 

concentration resulting from a temperature inversion during RaLa Run No. 21 showed a mixture of 

131I, 133I, and 135I (Burnett 1947a). A description of an investigation of elevated airborne activity in July 

1947 includes the assumptions used in computing airborne concentration from the count rate from the 

CAM filter. These assumptions include a collection efficiency of 100% for the activity, which from 

decay curves appeared to be 131I, and was assumed to be 131I for comparison against tolerance levels 

(Burnett 1947b). There are several instances in the late 1940s where the CAMs “blocked out,” 

meaning they went off-scale for a period. These instances typically occurred during periods of RaLa 

dissolvings, when radioiodines would have been a principal constituent of the activity. However, 

instances of off-scale activity would occur during RaLa runs after dissolving operations were 

completed, suggesting ground-level particulate emissions from waste jetting and sparging operations 

could peg the monitors, though radioiodines might have been the significant contributor in these 

instances as well. Other isotopes associated with short-duration instances of increased air activity 

include 140Ba (in January 1948) (Burnett 1948). In some cases, upsurges in air activity associated 

with temperature inversions could not be associated with a specific Laboratory process, and the 

activity decayed rapidly. Cases of increased air concentrations accompanying temperature inversions 

that were primarily 41Ar activity are described in (Morgan 1947). Several short-duration instances of 

elevated air activity are mentioned in the waste monitoring group reports from the late 1940s, albeit in 

a nonquantitative manner. The reports note instances of increased air activity if the measured value 

exceeded 10% of the 131I tolerance level for that era of 8.5 × 10-8 μCi cc-1. Note: Reliability of the 

CAMs during this period was an issue, during 1947 in particular, with data loss rates as high as 56% 

noted. Loss rates appear to have been significantly reduced by mid-1948. 

The airborne radioparticulate concentration results determined from the three CAM networks were 

primarily driven by fallout from above-ground nuclear weapons testing between 1960 and 1965. 

Fallout affected the concentration data for other periods as well, but between 1960 and 1965 

contributions from fallout to airborne gross beta concentrations reported by the ORNL LAM network 

dominated those from site operations. Fallout from a number of Soviet tests beginning in September 

1961 caused average concentration levels as determined by the LAM and RAM stations to increase 

by approximately 2 orders of magnitude for the third quarter of 1961 (Hart 1961b). Fourth quarter 

1961 data were similarly affected, although data for September represented the peak for CAM results. 

(The results from gummed-paper fallout trays did not peak until November of that year.) LAM and 

RAM data for September through December 1961 emphasize the magnitude of the effect of fallout on 

airborne concentration measurements, as the average concentration from the 8 RAM stations 

exceeded that of the 10 LAM stations. Thus, the LAMs were effectively blinded to airborne 

radioactivity originating from ORNL during that time. The same arguments hold true for particulate 

data from autoradiographs of the CAM filters and from material collected on the gummed-paper fallout 
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 24 of 41

trays. Gamma spectrometry performed on fallout material from Soviet tests collected in September 

1961 showed a combination of 237Np and mixed fission products, with the latter consisting primarily of 

140Ba/La, 132Te/I, 99Mo, 95Zr/Nb, and 131I (Hart 1961b). 

Figure 4-1 is a plot of annual average gross beta concentrations for the LAM and RAM networks for 

1956 through 1983. The effect of fallout on LAM data between 1960 and 1965 is apparent. 

0.00E+00

1.00E-12

2.00E-12

3.00E-12

4.00E-12

5.00E-12

6.00E-12

7.00E-12

8.00E-12

9.00E-12

1955 1960 1965 1970 1975 1980 1985

Gross Beta Concentration (μCi/cc)

LAMs RAMs

Figure 4-1. Annual average gross beta concentration data from LAM and 

RAM networks for 1956 through 1983. 

The fact that gross beta concentration results for the ORNL LAM network consist of little more than 

fallout and 7

Be from 1960 on makes the matter of assigning a nuclide mix for dose assessment 

somewhat arbitrary. The true net inhalation dose an individual might have received from material 

originating from ORNL would be only a small fraction of that assigned using the gross beta 

concentration. Further, the dose associated with gross measurements would be minimal regardless

of the nuclides assumed, given the low concentrations. The same is true for the early years of 

Laboratory history; however, for this period most of the activity collected on the filters originated from 

Laboratory operations. Much is known about nuclides that would have been present in the gross beta 

activity collected during this period; thus, a representative, albeit conservative, nuclide identification 

can be assigned to gross beta results prior to 1960 for dose assessment. 

The principal contributors to airborne gross beta activity collected by the LAM stations in the early 

period of Laboratory history would have been emissions associated with the RaLa process. RaLa 

operations were carried out at ORNL from 1944 to 1956. On an activity basis, the most abundant 

radionuclides present in RaLa emissions (excluding noble gases and their decay products) would 

have been isotopes of iodine, ruthenium, cerium, barium, lanthanum, zirconium, and niobium. Iodine 

and ruthenium were the more volatile of these, and thus made up the majority of the activity collected 

in periods of RaLa operations (RaLa was a batch process carried out several times a year). While the 

particulate filters in the LAMs would have collected some of the airborne radioiodines present, the 

highly variable and uncertain nature of the retention efficiency precludes making a defendable 

estimation of how much. Thus, while the gross beta results included some contribution from 

radioiodines (primarily 131I), the uncertainty associated with this contribution is such that it makes 

subtracting it impractical. Subtracting the 131I contribution to gross beta results would be desirable 

because 131I concentrations are treated separately in this document. However, including them with 
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gross beta results is conservative and, in reality, of little consequence given the small magnitude of 

the concentrations. 

Therefore, in light of the above, if it is desirable to assign a dose contribution from the gross beta 

(mixed fission product) concentration data provided in Attachment 4B, it is recommended the activity 

be treated as either 106Ru or 144Ce, depending on the organ of interest. Both should be assumed to be 

absorption type S. Along with 131I, the principal constituent of the particulate activity collected on LAM 

filters in the RaLa years was probably ruthenium (in a mixture of 103Ru and 106Ru); and Class S 106Ru 

is the bounding radionuclide of those that might have originated from ORNL. However, there may be 

cases (organs) where 106Ru is not bounding relative to 144Ce, in which case the latter should be 

selected. While it is true that isotopic analyses performed on LAM filters in later years did sometimes 

show the presence of 90Sr, it is not thought 90Sr would ever have been a significant component of the 

gross beta activity that actually originated from ORNL. Treating gross beta results as 106Ru or 144Ce 

should bound any real contributions to a worker’s dose from ORNL emissions after 1960 since very 

little of the total activity actually originated at the Laboratory. With all that said, however, dose 

reconstructors might find it easier to simply ignore gross beta (mixed fission product) intakes 

altogether since they likely will be only a minute fraction of the dose from other contributors (i.e., 131I or 

tritium). This is especially true given the conservative nature of the concentration data provided for 

131I and tritium in Attachment 4B. 

The gross beta (mixed fission product) concentration data in Attachment 4B are site-wide averages 

reported by the Laboratory. Specifically, they are the average of results from the stations in operation 

around the main ORNL campus plus, in later years, those for the HFIR area and the Low-Level 

Counting Laboratory about 2.5 miles east of the main complex. The results do not include 

contributions from outlying areas that could have brought down the average. As with the onsite 

exposure rate data described elsewhere in this document, use of a site-wide average is appropriate, 

given it would seem to be the likely outdoor exposure profile for an individual working at or visiting the 

ORNL site. However, as discussed above, contributions to a claimant’s dose from inhalation of mixed 

fission products in the outdoor air should be negligible in relation to that from 131I or tritium. This is 

especially true given that results after 1960 include very little contribution from materials originating 

from ORNL. 

4.4.2

A study was performed in the early 1960s to assess the stack sampling system for the 3018 stack visà-vis sample line losses (Youngblood and Fish 1961). At this time, the 3018 stack served both the 

ORNL Graphite Reactor and the Low-Intensity Test Reactor (LITR). The investigators determined 

that a quantitative estimate of the sampling error for particulates could not be made due to the lack of 

information on particle size distributions and the proximity of the sampling nozzle to flow disturbances. 

They estimated the loss of iodine vapor in the system to be at least 30%, but could not perform a 

more precise estimation due to turbulent flow conditions. They did assert that “sample losses are 

expected to be considerable in the present system” (Youngblood and Fish 1961). 

Radioactive Airborne Emissions

Youngblood and Fish (1961) did not directly evaluate the sampling system for the principal ORNL 

stack (3039), but did mention the fact that the filter for the 3039 monitoring system only showed a 

trace of 106Ru following the November 1959 fallout incident. In contrast, the sample line upstream of 

the filter is reported to have contained enough deposited material to give a direct reading of 10 mrad 

hr-1. The authors cite this as evidence that the sampling system for the 3039 stack was insensitive to 

the release of larger particles from agglomeration on interior surfaces of the fan and duct systems 

(Youngblood and Fish 1961). 
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ORNL began reporting noble gas releases to the AEC in 1970 in terms of annual totals (in curies) of 

85Kr and 133Xe. The quantities (activities) reported were determined via a conservatively selected ratio 

method for demonstrating compliance with contemporary MPC values. ORNL felt that these totals 

represented a gross upper bound and that actual releases were probably an order of magnitude lower 

(McVey 1972). ORNL used the conservative ratio method to assert noble gas releases because it did 

not have the capability to monitor the individual streams feeding to the central offgas system and the 

nature of the various processes that generated noble gas effluents was such that the nuclide mix 

varied widely (McVey 1972). Thus, noble gas releases asserted by ORNL appear to be conservative 

by a significant margin. 

ORNL did not begin reporting estimates of airborne tritium releases until 1972, and did not begin 

monitoring for tritium releases until 1984. The release estimates for 1972 to 1983 were based on 

throughput (i.e., the amount of tritium handled). The most significant airborne tritium releases in 

ORNL history occurred from approximately 1980 through 1991 (see Attachment 4A). 

ORNL began reporting annual releases of 131I in 1961 and continues to do so. Although the problems 

and issues associated with sampling for radioiodines (i.e., accounting for the large differences in 

behavior and reactivity among the different chemical species present in an effluent stream) were 

beginning to be understood around that time, it was nonetheless felt the releases reported for that era 

should be adjusted upward to account for potential losses in the sampling system and to ensure 

favorability to the claimant. 

Organic iodides are carried through long sample lines with little loss. However, reactive iodine (I2) can 

suffer from significant line losses, though some of the I2 that deposits on the walls is later 

resuspended. Because the true mix of iodine species that were present in the 3039 stack exhaust 

cannot be known, a factor of 2 was applied to the 131I releases asserted by ORNL for 1961 through 

1965. The asserted release data were not needed to compute onsite airborne concentrations for 

1966 through 1984 because representative monitoring data were available. A correction was not 

applied to the asserted releases after 1984 (when monitoring data are no longer available) because 

the releases reported at that time should meet modern standards, though they are minimal in terms of 

dose consequence. 

Although it was chosen somewhat heuristically, the factor of 2 applied to the asserted 131I release data 

for 1961 through 1965 should be reasonably accurate in light of what is known about sample line loss 

mechanisms for reactive iodine and the fact that releases in this period were small in terms of dose 

consequence. 

4.4.3

The instruments and gamma dosimeters used to perform exposure rate measurements on the ORNL 

site over its history were originally calibrated to 226Ra. Between 1970 and 1976, the method used to 

measure onsite exposure rates at ORNL switched from using film badges to using TLDs. TLDs were 

in use at the PAM stations to measure environmental exposure rates as early as 1973. At some point 

the photon source used to calibrate the dosimeters changed from 226Ra to 137Cs; however, for this 

document all exposure rate measurements have been treated as equivalent regardless of the 

calibration spectrum used or the reporting units (some measurements are reported in rad rather than 

R). 

Onsite Exposure Rate Measurements

Given the number of measurements taken, the fact that they reflect known off-normal releases of 

radioactive material at the ORNL site and compare favorably with continuous monitoring practices 

once these were initiated, it has been assumed that periodic exposure rate measurements performed 
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using portable instrumentation adequately represented ambient conditions at the times they were 

performed. 

All site-wide annual average exposure rate values provided in this document include natural 

background and contributions from non-ORNL-related activities, such as fallout. Fallout from 

atmospheric nuclear weapons testing was observed to have a minor effect on local exposure rate 

measurements on the ORNL site. Significant fallout (such as that from Soviet weapons tests 

beginning in September 1961) was observed to drive local ambient exposure rates higher, but only 

slightly so in relation to local sources. Hence, while fallout did have a positive bias on local exposure 

rate measurements, the magnitude of this bias does not ever appear to have been larger than around 

10% or so. For comparison, fallout from 1961 Soviet weapons tests raised background levels at 

remote locations by around 30% (Hart 1962). 

Annual average site-wide exposure rate data are not available for 1944 through 1946. The annual 

site-wide average determined for 1947 was therefore assigned for 1944 to 1946 as well. The largest 

contributor to elevated local exposure rates during that time was the RaLa process, from exposure 

rates around the process building itself while runs were in progress and from high-level liquid wastes 

eventually discharged to the settling basin. Data from the X-22 chambers from 1944 and 1945 show 

noble gas emissions to be only a minor contributor to onsite exposure rates. The number of shortdecayed fuel slugs from the Graphite Reactor used in the RaLa program in 1944 and 1945 was much 

less than that for 1946, and the amount of material processed in 1946 was comparable to that 

processed in 1947 [see ChemRisk (1999) for details]. Thus, applying the 1947 exposure rate for the 

prior years should be claimant-favorable. 

No onsite exposure rate data have been located to date for the LAM stations for 1971 through 1975. 

Annual average site-wide exposure rate data began being reported again in 1976, by which time 

dosimeters deployed at the LAM locations had changed from film badges to TLDs. To fill the period of 

missing data, the annual site-wide average exposure rate measured for 1970 was assumed for 1971 

through 1975. The result for 1970 was 0.072 mR hr-1. In 1976, when the annual average site-wide 

exposure rate was again reported, the value was 0.056 mR hr-1. 

No annual average exposure rate data for the ORNL site have been located for any years after 1984. 

The 1984 average was therefore assumed for 1985 through 2003. The same value was applied for 

1982 because a quantitative measure of the average exposure rate was lacking for that year as well. 

The annual average exposure rate for the ORNL site did not show appreciable change from 1979 

through 1984. 

4.5 DERIVATION OF ONSITE AIRBORNE CONCENTRATIONS FOR RADIONUCLIDES OF 

CONCERN

The three principal contributors to inhalation dose individuals could have been exposed to in the 

outdoor air on the ORNL site were 131I, particulate mixed fission products (MFPs), and tritium. Lack of 

available monitoring data, for 131I in particular, required that bounding values of dispersion coefficients 

coupled with available source term estimations be used to establish the needed onsite airborne 

concentration and intake data for certain periods. The dispersion coefficients used for these 

calculations are discussed in the next subsection, followed by discussion of the derivation of the 

airborne concentration data for outdoor air at ORNL for the three principal contributors to onsite dose. 

All concentration data are given in terms of annual averages.
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Attachment 4B to this document provides a summary of the annual average airborne concentration 

data and the associated intakes for an individual assumed to be exposed to the average 

concentration for 2,000 hours breathing at a rate of 1.2 cubic meters per hour. 

4.5.1

The lack of reliable measurements of airborne tritium and 131I concentrations at ORNL meant an 

alternative method had to be used to derive estimates of these values. Available source term 

information (if available) or release estimates, coupled with dispersion coefficients were therefore 

used. However, given the small size of the ORNL site, bounding values for dispersion coefficients 

were used for elevated release points. A bounding value was used for low-elevation releases such as 

fugitive emissions directly from buildings and short, local stacks, as appropriate. 

Dispersion Coefficients for Principal Elevated Release Points

Dispersion data asserted for the more important ORNL stacks are summarized in Table 4-2. These 

values were used as needed to provide estimates of local onsite airborne concentrations of tritium and 

131I for periods when reliable concentration measurements were not available. 

Table 4-2. Dispersion coefficients used for ORNL airborne release points. 

Stack Factor Dimensions Reference Remarks

3039 (formerly 900) 2.226 × 10-7 sec m-3 (Binford, Hamrick, 

and Cope 1970)

Maximum, annual-average χ/Q 

value; occurs 2,000 meters 

northeast of stack. 

3020 (formerly 205) 2.615 × 10-7 sec m-3 (Binford, Hamrick, 

and Cope 1970)

Maximum, annual-average χ/Q 

value; occurs 2,000 meters 

northeast of stack. 

Roof level (50 feet) 2.1 × 10-4 sec m-3 (Browning 1962) Assumes worst-case 

meteorological conditions

In a report prepared by ORNL for public hearings on radioactive waste disposal held before the Joint 

Congressional Committee on Atomic Energy in early 1959, reference is made to an unpublished study 

performed by the U. S. Weather Bureau by R. E. Meyers and D. R. Purdy titled “Calculations for Unit 

Emissions of Air-borne Contaminants.” According to Browder (1959), this study included calculations 

of maximum ground concentrations per unit discharge rate for varying weather conditions for the 

ORNL stacks, and asserted a probable minimum dilution factor of 10-3 for each of the three stacks. 

This dilution factor was the basis for the ORNL stack monitoring program in the early 1960s, or more 

specifically, was the basis for calculations of maximum allowable discharge rates. A determination of 

the maximum allowable release of 131I via rooftop hood exhausts from 1962 included calculation of a 

worst-case dispersion coefficient for releases 50 feet above the ground. The dispersion coefficient 

computed for a roof-level (50-foot) release under a night-time temperature inversion and 5-mph winds 

was 2.1 × 10-4 sec m-3 (Browning 1962). This value was used to estimate ground-level concentrations 

resulting from local, low-elevation releases on the ORNL site. 

4.5.2

The largest airborne releases of 131I were associated with RaLa operations, which took place at ORNL 

between 1944 and 1956. Processing of short-decay irradiated fuel in the Thorex process in 1956 and 

1957 was the next largest source of airborne 131I emissions, followed by the production of 131I for 

distribution under the ORNL isotopes program. ORNL did not start reporting airborne releases of 131I 

until 1961, by which time releases had diminished considerably. Similarly, the Laboratory did not start 

monitoring onsite airborne concentrations of 131I until 1966, when it added charcoal cartridges to the 

sampling chain of the LAM stations. The Laboratory maintained an air monitoring program for 

airborne particulate radioactivity from early in its history, but the data cannot be used to establish 

Onsite Airborne Concentrations of 131I
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 29 of 41

reliable estimates of local airborne 131I concentrations. Particulate filters are in general a poor 

collector for radioiodines. They will collect iodine that has sorbed onto particles, plus a small fraction 

of the reactive I2 species, but the retention efficiency for the reactive species is highly variable and a 

function of the incident concentration and the relative humidity, among other things. Thus, given the 

fact that neither the retention efficiency of the particulate filters used in the LAMs nor the overall mix of 

the iodine species in the local air at ORNL can be known, no defendable assertions of airborne 

radioiodine concentrations can be drawn from particulate monitoring results. Thus, for 1944 through 

1965, airborne 131I concentrations on the ORNL site had to be established using information on 

emission rates and bounding values of dispersion coefficients. 

For 1944 through 1956, the principal source of airborne 131I emissions at ORNL was the RaLa 

program. 131I releases associated with RaLa processing at ORNL were studied in detail as an 

element of the Oak Ridge Dose Reconstruction and are documented in ChemRisk (1999). Iodine-131 

releases from the RaLa program are documented in Chapter 3 of this reference in terms of annual 

total emissions. Separate source terms are provided for the stacks used to ventilate dissolver 

offgases and for the process building itself. This source term information was used to derive local 

onsite concentrations of 131I needed for this document by converting annual releases to release rate 

and then multiplying by bounding dispersion coefficients for either the stack or roof-level emissions, as 

appropriate. The dispersion coefficients used were as follows:

• 2.62 × 10-7 seconds per cubic meter for the 3020 stack, which ventilated dissolver offgases 

from 1944 until 1950

• 2.23 × 10-7 seconds per cubic meter for the 3039 stack, which ventilated dissolver offgases 

from 1950 until 1956

• 2.1 × 10-4 seconds per cubic meter for direct (fugitive) emissions

All of the dispersion coefficients used are bounding values, as described in Section 4.5.1. The results 

of the calculations of bounding, local 131I concentrations at ORNL determined from the RaLa source 

term are summarized in Table 4-3. Note these values include releases associated with the April 1954 

accident, as documented in Section 3.15 of ChemRisk (1999). 

Table 4-3. Bounding values of onsite airborne 131I concentrations at ORNL from RaLa processing. 

Annual Average Annual Average Total

Concentration Concentration Annual Average

Total Released Total Released from Stack from Building Concentration

Year from Stack (Ci) from Building (Ci) (μCi/cc) (μCi/cc) (μCi/cc)

1944 262.1 0.5 2.2E-12 3.5E-12 5.7E-12

1945 1371.4 62.4 1.1E-11 4.2E-10 4.3E-10

1946 830.1 16.1 6.9E-12 1.1E-10 1.1E-10

1947 1090.1 19.1 9.1E-12 1.3E-10 1.4E-10

1948 640.9 16.1 5.3E-12 1.1E-10 1.1E-10

1949 953.1 12.1 7.9E-12 8.0E-11 8.8E-11

1950 500.1 5.3 3.5E-12 3.6E-11 3.9E-11

1951 571.0 2.6 4.0E-12 1.7E-11 2.1E-11

1952 2010.3 17.1 1.4E-11 1.1E-10 1.3E-10

1953 3290.4 40.2 2.3E-11 2.7E-10 2.9E-10

1954* 3546.8 305.0 2.5E-11 2.0E-09 2.1E-09

1955 2330.2 28.2 1.6E-11 1.9E-10 2.0E-10

1956 3370.3 26.1 2.4E-11 1.7E-10 2.0E-10

*includes totals from April, 1954 accident.
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 30 of 41

After RaLa processing at ORNL was terminated in October 1956, the principal source of 131I 

emissions from the site appears to have been the short-decay Thorex runs that took place in the 3019 

Building between July 1956 and November 1957. The total quantity of 131I available for release from 

these runs was estimated by ChemRisk (1993a) to be 906 Ci for 1956 and 14,600 Ci in 1957. These 

estimates were used to determine bounding, ground-level concentrations at the ORNL site using 

essentially the same approach as that used for RaLa releases. The total 131I inventory for each year 

was multiplied by a factor of 0.027 to provide a total released. The factor of 0.027 is the ratio of 131I 

releases across all RaLa runs to the available inventory. The assumption that releases from the 

Thorex and RaLa processes were comparable is based on the commonalities between the process 

equipment used for each, despite the fact Thorex was conducted in the Pilot Plant (Building 3019). 

The assumption of similar release characteristics for 131I between the two processes is supported by 

interviews with individuals responsible for the design of the offgas treatment systems for the Thorex 

plant conducted as an element of the Oak Ridge Dose Reconstruction Project. (For example, see the 

interview with O. Yarbro documented as Accession Number 2562 in the Oak Ridge Dose

Reconstruction Database, which is an element of the DOE Comprehensive Epidemiologic Data 

Resource (CEDR) hosted on the Internet by Lawrence Berkeley National Laboratory.) The total 

releases computed using the factor of 0.027 were partitioned between the fraction released directly 

from the 3019 Building and that released via the 3020 stack by applying a ratio of 0.01 (i.e., 1% of the 

total was assumed to be released directly, and the remainder via the stack). The 1% value was 

selected based on the RaLa source term, as documented in (ChemRisk 1999). The resulting total 

releases for the two pathways were then converted to ground-level concentrations in the same 

manner as that used for the RaLa releases. Calculations of the bounding ground-level 131I 

concentrations associated with the short-decay Thorex runs of 1956 and 1957 are summarized in 

Table 4-4. 

Table 4-4. Bounding values of onsite airborne 131I concentrations at ORNL from short-decay Thorex 

processing. 

Annual Average Annual Average Total

Concentration Concentration Annual Average

Total Released Total Released from Stack from Building Concentration

Year from Stack (Ci) from Building (Ci) (μCi/cc) (μCi/cc) (μCi/cc)

1956 24.2 0.2 2.0E-13 1.6E-12 1.8E-12

1957 390.3 3.9 3.2E-12 2.6E-11 2.9E-11

Neither local concentration measurements nor source term estimates for 131I are available for 1958 to 

1960. The only data located to date for this period are isotopic sales summaries for 1958, in which 

the Laboratory reported 131I production in the amount of 4,040 Ci and sales of 1,074 Ci (ChemRisk 

1993b). The difference between the quantity produced and that sold is attributed to radioactive 

decay. No isotopic sales data have been located for 1959 and 1960, although the program was still in 

operation at that time. Thus, assumptions were made based on the growth in 131I sales. From sales 

data presented in Table 2-6 of ChemRisk (1993b), it appears that 131I sales experienced fairly 

consistent growth between 1955 and 1958. In fact, the growth between 1956 and 1957 and between 

1957 and 1958 are identical at 21%; thus, this factor was used to project 131I production for 1959 and 

1960. Doing so, one gets projected 131I production values of 4,889 Ci and 5,915 Ci for 1959 and 

1960, respectively. 

The 131I production values for 1958 to 1960 were converted to annual releases by applying the factor 

of 0.027 determined for RaLa releases, knowing this should be conservative for Laboratory facilities 

used for 131I production. Doing so yields average release estimates of 109 Ci, 132 Ci, and 160 Ci for 

1958, 1959, and 1960, respectively. These values compare favorably with 131I releases reported by 
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 31 of 41

ORNL for subsequent years. Annual average ground-level concentrations were computed by 

applying the bounding dispersion coefficient for the 3039 stack (2.23 × 10-7 sec m-3). 

As discussed in Section 4.4.2, a factor of 2 was applied to 131I release data reported by ORNL for 

1961 through 1965 to account for sample line loss. The doubled release estimates were used to 

compute ground-level concentration by applying the bounding dispersion coefficient for the 3039 

stack. 

The average values reported by the Laboratory from charcoal cartridge measurements from the LAM 

stations were used for 1966 through 1984, when these data ceased being reported. The annual 

average concentration for 1984 was assigned to 1985 through 2003 for completeness. 

The 131I concentration data described above are summarized in Attachment 4B. It has been assumed 

that 131I emissions from isotope production activities from 1944 through 1957 are negligible in relation 

to bounding assessments performed using source term information for the RaLa and Thorex 

programs. For dose assessment, it is recommended the airborne 131I be considered to be in a soluble 

or reactive (vapor) form. 

4.5.3

The concentration and intake data for mixed fission products in Attachment 4B are reported annual 

averages for the ORNL LAM network. These data are gross beta measurements and include 

contributions from fallout and other sources, such as the 7

Be naturally present in the atmosphere. 

These contributions are the dominant constituent of the gross beta activity measured after 1960. The 

concentration values are such that the dose contribution from airborne mixed fission products should 

at all times be negligible in relation to that from 131I or tritium. However, if it is desired to include the 

dose contribution from mixed fission products, it is recommended the activity be assumed to be Type 

S 106Ru or 144Ce, depending on which yields a higher dose for the organ of interest. Gross beta 

(mixed fission product) concentration results are described in more detail in Section 4.4.1. Note: The

average concentration value for 1949 was assumed for 1944 through 1948. 

Onsite Airborne Concentrations of Mixed Fission Products

4.5.4

There have been two principal sources of airborne tritium emissions from ORNL over its history: the 

3039 (central offgas system) stack and the local stack for the TTFF, Building 7025. The TTFF was 

not operational prior to 1967; hence, all airborne tritium releases prior to that time were via the 3039 

stack alone. The combination of the large dilution afforded by the 3039 stack and the small 

magnitude of releases during that time was such that airborne tritium emissions from ORNL prior to 

1967 are negligible in relation to the dose contribution from airborne 131I. Isotope sales of tritium in 

that period were too little to have resulted in significant onsite concentrations from airborne emissions. 

Ohnesorge (1986) provides the following isotopic sales information for tritium for the period prior to 

1972:

Onsite Airborne Concentrations of Tritium

• Total sales prior to 1965 were 363,000 Ci.

• Sales varied from 121,000 to 334,000 Ci from 1965 through 1971. 

• Sales for 1972 were 161,000 Ci, corresponding to a total reported release of 1800 Ci.

ORNL did not begin routinely reporting airborne tritium emissions until 1972; however, a March 1971 

letter summarizing liquid and gaseous releases from ORNL asserts airborne tritium releases of 4,500 

Ci and 300 Ci for 1969 and 1970, respectively (Sapirie 1971). 
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 32 of 41

The period of peak airborne releases of tritium from ORNL was from 1980 through 1991. Large-scale 

tritium operations at ORNL ceased in 1989, but releases persisted for some time thereafter from offgassing of tritium-bearing systems. By 1988, Building 3033 was the principal source of tritium 

emissions from ORNL, handling roughly 1,000,000 Ci per year in the purification of tritium received 

from the Savannah River Plant for redistribution to industrial and research users (Shor 1988). 

Airborne releases of tritium from Building 3033 were via the central (3039) stack. In addition, tritium 

was handled in Building 7025 in the manufacture of accelerator targets and material test samples. 

Quantities handled in Building 7025 were roughly 30,000 Ci per year around 1988 (Shor 1988). Thus, 

during its period of peak emissions, the quantity of tritium handled in Building 7025 was roughly 3% of 

that handled in Building 3033. 

To date, no monitoring data for airborne tritium concentrations on the ORNL site have been located. 

Concentrations were therefore established using a combination of release data and bounding 

dispersion coefficients, as was done to obtain 131I concentrations for the period prior to 1966. Onsite 

airborne tritium concentrations were established in this manner for 1967 to the present. Tritium 

releases were negligible in terms of onsite concentrations prior to operation of the TTFF, which 

released tritium to the atmosphere via a short local stack rather than the 250-foot central stack that 

served the 3033 Building. The fact that airborne tritium emissions from the TTFF were via its own 

local stack makes this facility the dominant source of onsite tritium concentrations at ORNL (under the 

conservative assumptions employed in this assessment). 

Tritium release data asserted by ORNL are available for 1969, 1970, and from 1972 through 2002 

(see Attachment 4A). Because the release estimations are for the most part based on material 

handling and inventory evaluations, no corrections for sampling errors were deemed necessary. 

Tritium release estimates were therefore needed for 1967, 1968, and 1971. From the annual tritium 

sales data provided by Ohnesorge (1986), the largest quantity of tritium sold from 1965 through 1971 

was 334,000 Ci. Ohnesorge (1986) also provides annual sales for 1972, which is the first year ORNL 

began routinely reporting tritium releases. The ratio of the reported release for 1972 (1,800 Ci) to 

annual sales (161,000 Ci) gives the factor 0.011, which is the value ORNL used to compute the 

release. Applying this factor to the largest quantity of tritium sold in the period of interest (334,000 Ci) 

gives a release of 3,674 Ci. This value was assigned as the amount released for 1967, 1968, and 

1971, implying the conservative assumption that the maximum sales occurred in all three of those 

years. 

Annual average airborne tritium concentrations for the ORNL site were computed using the annual 

release data described above. Concentrations were computed by assuming 97% of the total release 

was via the 3039 stack and the remainder was via the local 7025 stack. The dispersion coefficients 

used for these two release points were 2.23 × 10-7 sec m-3 and 2.1 × 10-4 sec m-3, respectively (see 

Section 4.5.1). The concentration computed for 2002 was assigned for 2003. 

The concentration values and associated intakes computed as described above are summarized in 

Attachment 4B. Note: The intake values include a factor of 1.5 applied to account for direct 

absorption through the skin. Dose reconstructors should assume the tritium is in the physical form of 

HTO (tritiated water vapor). A geometric standard deviation of 3 should be assumed for the intake 

values given in Attachment 4B. 

4.6 ONSITE EXPOSURE RATE DATA

Annual-average exposure rate data for the ORNL site are summarized in Attachment 4C for 1944 

through 2003. The methods used by ORNL to determine these results are discussed in Section 4.3.6. 
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 33 of 41

Assumptions associated with these data, including those used to fill gaps where measured values 

were not available, are discussed in Section 4.4.3. Two important points concerning these data are: 

• The exposure rates are gross values, i.e., background is included. 

• The exposure rate data are site-wide averages. 

The annual average exposure rate data in Attachment 4C of this report represent site-wide averages. 

“Site-wide” in this context means the average value reported by the Laboratory in the periodic reports 

of the Health Physics Division. These average values consisted only of measurements made in the 

core areas of the Laboratory and areas where exposure rates where known or suspected to be 

elevated. The values, therefore, do not suffer from low bias from having a disproportionate number of 

measurements at or near background. The site-wide average was selected as the measure of 

interest because it was felt to be the most representative of an unmonitored worker’s exposure profile. 

ORNL is a relatively small open campus where visitors and workers tend to move around during the 

time they are on the site. An individual’s exposure, therefore, would include a number of different 

sources and areas as he/she moved about. The site-wide average exposure rate data should 

therefore provide a defendable basis for assigning external environmental dose for unmonitored 

workers. Indeed, the site-wide averages include measurement locations where exposure rates were 

chronically much higher than those for the other measurement locations and thus served to drive the 

average higher. These locations include the vicinity of the 706-D fan house, which operated until 

about 1950, and the vicinity of the settling basin. Other locations would raise the site-wide average 

depending on what activities were taking place, such as locations in the vicinity of Buildings 706-A or 

706-C in the mid-to-late 1940s. The fact that these instances are included in the average exposure 

rate data means the site-wide averages should be sufficiently bounding for an unmonitored worker. 

The ORNL monitoring program kept up with the footprint of the Laboratory as its facilities expanded 

over the years, so site-wide averages should always represent the complement of areas of elevated 

exposure an individual might have encountered. 

A caveat associated with the use of site-wide average exposure rates is that these values might not 

be appropriate for cases in which a composite exposure profile for an unmonitored individual is 

inconsistent with a known or asserted exposure history. If a worker’s dose for a particular period has 

to be considered to have come from a specific source or location, the dose reconstructor will have to 

make a judgment about the magnitude of the site-wide average in relation to the nominal exposure 

rate for the specific source or location. However, the site-wide average values in this document 

should be higher than exposure rates that were present on most of the ORNL site because they 

include locations where local exposure rates were significantly elevated. For example, exposure rate 

measurements for a location 15 feet from the southwest corner of the settling basin for 9 of the 13 

months from September 1948 through September 1949 average 3.5 mR hr-1, and range from 1.7 to 

8.5 mR hr-1 (in August 1949). Similarly, eight measurements at a location 10 feet north of the 706-D 

(RaLa Building) fan house for the same period average 9.4 mR hr-1 and range from 2.8 to 29 mR hr-1. 

In comparison, the annual averages for the Laboratory grounds for 1948 and 1949 were 0.29 and 

0.46 mR hr-1, respectively. These are the two highest averages recorded for the Laboratory over its 

history, and are based on periodic (approximately monthly) measurements performed at 45 specific 

locations on the Laboratory grounds, including those in the vicinity of the settling basin and the 706-D 

fan house. Note. The 706-D fan house was not used after 1949 [i.e., after the RaLa Building was tied 

into the (then new) central offgas system]. Also note the substantial drop in site-wide averages after 

the 706-D fan house was retired (see Attachment 4C). 

Differences between site-wide average exposure rates and the exposure rate for a specific location 

would probably have been the largest earlier in the Laboratory’s history. This is when such 
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 34 of 41

differences would have been the most important as far as an unmonitored external dose a claimant 

might have received. 

Figure 4-2 is a plot showing the trend of the annual average exposure rates for the ORNL site. Note: 

The periods 1944 to 1946, 1971 to 1975, 1982, and 1985 to 2003 reflect assumed rather than 

measured values. 

0

0.05

0.1

0.15

0.2

0.25

0.3

0.35

0.4

0.45

0.5

1940 1950 1960 1970 1980 1990 2000

A

v

e

r

a

g

e

E

x

p

o

s

u

r

e

R

a

t

e

(

m

R

/

h

r

)

Year

Figure 4-2. Annual average exposure rates for ORNL. 
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 35 of 41

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Ohnesorge, W. F., 1986, Historical Releases of Radioactivity to the Environment from ORNL, 

ORNL/M-135, Oak Ridge National Laboratory, Oak Ridge, Tennessee, May. 

ORNL (Oak Ridge National Laboratory), 1983, Industrial Safety and Applied Health Physics Division 

Annual Report for 1982, ORNL-5962, Oak Ridge, Tennessee, December. 
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 38 of 41

Parker, H. M., 1944, Review of Air Monitoring Procedures at Clinton Laboratories, CH-2562, Oak 

Ridge National Laboratory, Oak Ridge, Tennessee, September. 

Sapirie, S. R., 1971, “Radioactive Effluents and Releases”, letter to R. F. Hibbs, U. S. Atomic Energy 

Commission, Oak Ridge Operations, Oak Ridge, Tennessee, March 16.

Shor, J. T., 1988, Tritium Effluent Reduction at Oak Ridge National Laboratory, ORNL/TM-10664, Oak 

Ridge National Laboratory, Oak Ridge, Tennessee, November.

Unknown, 1948, The Particle Problem at ORNL: An Historical Summary, ORNL Central Files Number 

49-1-49, Oak Ridge National Laboratory, Oak Ridge, Tennessee, December 30.

Youngblood, J. W., and B. R. Fish, 1961, Studies of ORNL Stack Monitoring – I. Improvement of the 

3018 Stack Sampling System, ORNL Central Files Number 61-8-90, Oak Ridge National 

Laboratory, Oak Ridge, Tennessee, October 6.

Wirth, J. E., K. Z. Morgan, and H. J. Curtis, 1945, Health Division and Biological Section of the 

Research Division Report for the Month Ending February 15, 1945, CH-2587, Oak Ridge 

National Laboratory, Oak Ridge, Tennessee, February 27. 
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 39 of 41

ATTACHMENT 4A

ANNUAL AIRBORNE RELEASES REPORTED BY OAK RIDGE NATIONAL LABORATORY 

(CURIES)

Year 131I 3

H 85Kr 133Xe 41Ar 138Cs

1961 42 -- -- -- -- --

1962 121.2 -- -- -- -- --

1963 54 -- -- -- -- --

1964 84.5 -- -- -- -- --

1965 18.4 -- -- -- -- --

1966 15.79 -- -- -- -- --

1967 22.3 -- -- -- -- --

1968 10.38 -- -- -- -- --

1969 16.38 4500 -- -- -- --

1970 1.43 300 15000 75000 -- --

1971 3.46 -- 15000 71000 -- --

1972 1.7 1800 15400 64900 -- --

1973 2.18 9100 14000 68600 -- --

1974 1.97 555 20000 99200 -- --

1975 2.1 534 17700 87500 -- --

1976 1.25 6019 11500 54000 -- --

1977 1.37 2524 8606 42030 -- --

1978 1.7 2500 12000 59000 -- --

1979 0.3 5109 10500 51190 -- --

1980 0.22 14800 8800 42800 -- --

1981 0.5 11300 6700 32400 -- --

1982 0.13 19000 11700 57100 -- --

1983 0.05 22200 11900 57700 -- --

1984 0.1 33400 14900 72700 -- --

1985 0.09 20180 6623 32280 -- --

1986 0.035 31000 10600 51000 -- --

1987 0.02 44050 4700 22700 -- --

1988 0.056 21000 6300 31000 -- --

1989 0.03 28000 16000 79000 -- --

1990 0.02 12000 89000 1500 -- --

1991 0.046 16000 5400 3100 -- --

1992 0.14 3000 -- -- -- --

1993 0.05 410 -- -- -- --

1994 0.065 450 21 260 -- --

1995 0.16 300 -- -- -- --

1996 0.28 603 280 1.1 -- --

1997 0.055 148 110 0.0068 10000 --

1998 0.062 110 260 320 7950 --

1999 0.058 103 480 4.6 12500 --

2000 0.075 107 283 2.21 3620 --

2001 0.13 49 490 0.488 -- 1360

2002 0.09 86 314 5.88 -- 1590

Notes:

Missing data implies release is unknown. It does not imply there was no release. 

41Ar and 138Cs data are only given for years when these releases represented maximum collective dose. 131I release reported for 1961 is an estimation. The values for subsequent years are based on in-stack sampling.

Tritium releases prior to 1984 are estimates based on inventory.
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 40 of 41

ATTACHMENT 4B

ANNUAL AVERAGE AIRBORNE CONCENTRATION AND INTAKE DATA FOR IODINE-131, 

TRITIUM, AND PARTICULATE MIXED FISSION PRODUCTS AT ORNL

Year 131I Tritium MFP 131I Tritium MFP

1944 2.1E-01 -- 7.4E-01 5.0E+02 -- 1.8E+03

1945 1.6E+01 -- 7.4E-01 3.8E+04 -- 1.8E+03

1946 4.2E+00 -- 7.4E-01 1.0E+04 -- 1.8E+03

1947 5.0E+00 -- 7.4E-01 1.2E+04 -- 1.8E+03

1948 4.2E+00 -- 7.4E-01 1.0E+04 -- 1.8E+03

1949 3.3E+00 -- 7.4E-01 7.8E+03 -- 1.8E+03

1950 1.4E+00 -- 6.7E-01 3.5E+03 -- 1.6E+03

1951 7.9E-01 -- 6.3E-01 1.9E+03 -- 1.5E+03

1952 4.7E+00 -- 1.1E-01 1.1E+04 -- 2.5E+02

1953 1.1E+01 -- 1.7E-01 2.6E+04 -- 4.1E+02

1954 7.6E+01 -- 1.8E+00 1.8E+05 -- 4.2E+03

1955 7.6E+00 -- 1.8E-01 1.8E+04 -- 4.4E+02

1956 7.4E+00 -- 1.0E-01 1.8E+04 -- 2.5E+02

1957 1.1E+00 -- 3.0E-01 2.6E+03 -- 7.3E+02

1958 2.9E-02 -- 9.3E-02 6.8E+01 -- 2.2E+02

1959 3.5E-02 -- 1.4E-01 8.3E+01 -- 3.4E+02

1960 4.2E-02 -- 1.7E-02 1.0E+02 -- 4.0E+01

1961 2.2E-02 -- 5.9E-02 5.3E+01 -- 1.4E+02

1962 6.3E-02 -- 1.4E-01 1.5E+02 -- 3.3E+02

1963 2.8E-02 -- 1.8E-01 6.8E+01 -- 4.4E+02

1964 4.4E-02 -- 4.8E-02 1.1E+02 -- 1.2E+02

1965 9.6E-03 -- 1.0E-02 2.3E+01 -- 2.5E+01

1966 8.5E-03 -- 6.3E-03 2.0E+01 -- 1.5E+01

1967 1.6E-03 2.8E+01 8.1E-03 3.8E+00 1.0E+05 2.0E+01

1968 4.8E-03 2.8E+01 1.1E-02 1.2E+01 1.0E+05 2.6E+01

1969 1.1E-02 3.4E+01 9.6E-03 2.8E+01 1.2E+05 2.3E+01

1970 1.1E-03 2.3E+00 1.2E-02 2.7E+00 8.3E+03 2.9E+01

1971 1.5E-03 2.8E+01 1.6E-02 3.6E+00 1.0E+05 3.8E+01

1972 1.4E-03 1.4E+01 4.8E-03 3.3E+00 5.0E+04 1.2E+01

1973 1.7E-03 7.0E+01 2.4E-03 4.2E+00 2.5E+05 5.8E+00

1974 1.1E-03 4.2E+00 3.7E-03 2.6E+00 1.5E+04 8.9E+00

1975 9.3E-04 4.1E+00 2.1E-03 2.2E+00 1.5E+04 5.2E+00

1976 9.6E-04 4.6E+01 1.7E-03 2.3E+00 1.7E+05 4.1E+00

1977 9.3E-04 1.9E+01 2.3E-03 2.2E+00 6.9E+04 5.5E+00

1978 1.1E-03 1.9E+01 3.3E-03 2.8E+00 6.9E+04 7.9E+00

1979 7.8E-04 3.9E+01 1.6E-03 1.9E+00 1.4E+05 3.9E+00

1980 1.3E-04 1.1E+02 1.9E-03 3.0E-01 4.1E+05 4.6E+00

1981 1.3E-04 8.6E+01 3.1E-03 3.1E-01 3.1E+05 7.4E+00

1982 1.3E-04 1.5E+02 1.6E-03 3.0E-01 5.2E+05 3.8E+00

1983 1.9E-04 1.7E+02 1.9E-03 4.4E-01 6.1E+05 4.5E+00

1984 1.5E-04 2.6E+02 1.0E-03 3.6E-01 9.2E+05 2.4E+00

1985 1.5E-04 1.5E+02 1.0E-03 3.6E-01 5.6E+05 2.4E+00

1986 1.5E-04 2.4E+02 1.0E-03 3.6E-01 8.5E+05 2.4E+00

1987 1.5E-04 3.4E+02 1.0E-03 3.6E-01 1.2E+06 2.4E+00

1988 1.5E-04 1.6E+02 1.0E-03 3.6E-01 5.8E+05 2.4E+00

1989 1.5E-04 2.1E+02 1.0E-03 3.6E-01 7.7E+05 2.4E+00

1990 1.5E-04 9.2E+01 1.0E-03 3.6E-01 3.3E+05 2.4E+00

1991 1.5E-04 1.2E+02 1.0E-03 3.6E-01 4.4E+05 2.4E+00

1992 1.5E-04 2.3E+01 1.0E-03 3.6E-01 8.3E+04 2.4E+00

1993 1.5E-04 3.1E+00 1.0E-03 3.6E-01 1.1E+04 2.4E+00

1994 1.5E-04 3.4E+00 1.0E-03 3.6E-01 1.2E+04 2.4E+00

1995 1.5E-04 2.3E+00 1.0E-03 3.6E-01 8.3E+03 2.4E+00

1996 1.5E-04 4.6E+00 1.0E-03 3.6E-01 1.7E+04 2.4E+00

1997 1.5E-04 1.1E+00 1.0E-03 3.6E-01 4.1E+03 2.4E+00

1998 1.5E-04 8.4E-01 1.0E-03 3.6E-01 3.0E+03 2.4E+00

1999 1.5E-04 7.9E-01 1.0E-03 3.6E-01 2.8E+03 2.4E+00

2000 1.5E-04 8.2E-01 1.0E-03 3.6E-01 2.9E+03 2.4E+00

2001 1.5E-04 3.7E-01 1.0E-03 3.6E-01 1.3E+03 2.4E+00

2002 1.5E-04 6.6E-01 1.0E-03 3.6E-01 2.4E+03 2.4E+00

2003 1.5E-04 6.6E-01 1.0E-03 3.6E-01 2.4E+03 2.4E+00

Notes:

Intake data assume inhalation rate of 2400 m3 per year. 131I should be assumed to be in vapor form. 

Tritium releases were not significant in terms of dose prior to 1967.

Tritium should be assumed to be in the form HTO.

Tritium intake values include a factor of 1.5 for skin absorption.

Dose from mixed fission products (MFPs) may be negligible relative to other contributors.

MFP concentrations included for completeness. Data are site-wide averages. 

Assume MFP concentrations are Type S 106Ru or 144Ce if dose assessment is desired. 

Assume geometric standard deviation of 3 for all values. 

Annual Average Concentration (Bq m-3) Annual Intake (Bq)
Effective Date: 05/07/2004 Revision No. 00 Document No. ORAUT-TKBS-0012-4 Page 41 of 41

ATTACHMENT 4C

ANNUAL AVERAGE SITE-WIDE EXPOSURE RATES FOR OAK RIDGE NATIONAL 

LABORATORY

Average

Exposure

Year Rate (mR/hr)

1944 0.14

1945 0.14

1946 0.14

1947 0.14

1948 0.29

1949 0.46

1950 0.08

1951 0.11

1952 0.28

1953 0.21

1954 0.14

1955 0.12

1956 0.14

1957 0.11

1958 0.13

1959 0.19

1960 0.13

1961 0.1

1962 0.106

1963 0.104

1964 0.081

1965 0.072

1966 0.075

1967 0.071

1968 0.076

1969 0.084

1970 0.072

1971 0.07

1972 0.07

1973 0.07

1974 0.07

1975 0.07

1976 0.056

1977 0.042

1978 0.037

1979 0.033

1980 0.032

1981 0.029

1982 0.03

1983 0.029

1984 0.031

1985 0.03

1986 0.03

1987 0.03

1988 0.03

1989 0.03

1990 0.03

1991 0.03

1992 0.03

1993 0.03

1994 0.03

1995 0.03

1996 0.03

1997 0.03

1998 0.03

1999 0.03

2000 0.03

2001 0.03

2002 0.03

2003 0.03

Notes: Precision of the results reflects that reported by ORNL, except where values were assumed 

based on results for other years. GSD of 3.0 should be applied to these values.