ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 27

Document Title:

Lawrence Livermore National Laboratory 

– Occupational Environmental Dose

Document Number: ORAUT-TKBS-0035-4 

Revision: 00

Effective Date: 11/18/2005

Type of Document: TBD

Supersedes: None 

Subject Expert: Billy R. Thomas

Document Owner

Approval: Signature on File Approval Date: 10/26/2005

Jay J. Maisler, TBD Team Leader

Approval: Signature on File Approval Date: 10/25/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 10/29/2005

Richard E. Toohey, Project Director

Approval: Signature on File Approval Date: 11/18/2005

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0035-4 Revision No. 00 Effective Date: 11/18/2005 Page 2 of 27

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

11/18/2005 00 New technical basis document for the Lawrence Livermore National 

Laboratory – Occupational Environmental Dose. Training required: 

As determined by the Task Manager. Initiated by Jay J. Maisler.
Document No. ORAUT-TKBS-0035-4 Revision No. 00 Effective Date: 11/18/2005 Page 3 of 27

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

4.0 Occupational Environmental Dose ........................................................................................... 6

4.1 Introduction .............................................................................................................................. 7

4.2 Internal Dose from Onsite Atmospheric Radionuclide Concentration ....................................... 7

4.2.1 Ambient Air Sample Collection Network ...................................................................... 7

 4.2.1.1 Pre 1971....................................................................................................... 7

 4.2.1.2 Post 1971 ..................................................................................................... 8

 4.2.1.3 Sampling Locations ...................................................................................... 8

 4.2.1.4 Radionuclides of Significance..................................................................... 10

4.2.2 Methodology..............................................................................................................11

4.2.3 Estimation of Annual Intake from Airborne Radionuclides ......................................... 15

4.2.3.1 Potential Intakes Prior to 1961.................................................................... 15

 4.2.3.2 Potential Intakes 1961 to 1971 ................................................................... 16

 4.2.3.3 Potential Intakes post 1971......................................................................... 16

4.3 Intake from Drinking Water.....................................................................................................17

4.4 Ambient Radiation..................................................................................................................19

4.5 Uncertainty.............................................................................................................................19

References...... ..................................................................................................................................24

Glossary.............................................................................................................................................27
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LIST OF TABLES

TABLE TITLE PAGE

4-1 Ambient air sampling station locations within LLNL boundary .................................................. 8

4-2 Ambient air sampling station locations at Site 300 ................................................................... 8

4-3a Maximum site-wide annual median intakes (Bq/yr) via inhalation........................................... 13

4-3b Maximum site-wide annual median intakes (Bq/yr) via inhalation........................................... 13

4-4a Site 300 maximum annual median intakes (Bq/yr) via inhalation............................................ 14

4-4b Site 300 maximum annual median intakes (Bq/yr) via inhalation............................................ 14

4-5 Maximum site-wide annual intakes (Bq/yr) via drinking water................................................. 18

4-6 Site 300 maximum annual intakes (Bq/yr) via drinking water.................................................. 18

4-7 External gamma radiation (mrem/yr)...................................................................................... 22

4-8 Site 300 external gamma radiation (mrem/yr)......................................................................... 23

LIST OF FIGURES

FIGURE TITLE PAGE

4-1 Air particulate and tritium sampling locations on Livermore site................................................ 9

4-2 Air particulate and tritium sampling locations at Site 300 and off site ..................................... 10

4-3 Calculated annual median concentrations of 239+240Pu, 1982-2001......................................... 17

4-4 Gamma dosimeter locations, Livermore Site, 2001 ................................................................ 20

4-5 Gamma dosimeter locations, Site 300 and vicinity, 2001........................................................ 21

4-6 Quarterly mean gamma dose measurements at the Livermore site 

 perimeter, 1988-2001............................................................................................................. 23
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ACRONYMS AND ABBREVIATIONS

CEDE committed effective dose equivalent

conc concentration

DOE U.S. Department of Energy

DU depleted uranium (less than 0.7 percent 235U)

EPA U.S. Environmental Protection Agency

LLNL Lawrence Livermore National Laboratory 

LRL Lawrence Radiation Laboratory

MEI maximally exposed individual

NESHAPS National Emissions Standards for Hazardous Air Pollutants

QA quality assurance

SW-MEI site-wide maximally exposed individual

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code
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4.0 OCCUPATIONAL ENVIRONMENTAL DOSE

Site Profiles, which include Technical basis documents (TBDs), are not official determinations made 

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

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

dose reconstructions at particular sites or categories of sites. They will be revised when additional 

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

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

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

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

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

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

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

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

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

operations ... pertaining to the Naval Nuclear Propulsion Program).” 42 U.S.C. § 7384l(12). 

Accordingly, except for the exclusion for the Naval Nuclear Propulsion Program noted above, any 

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

encompassed by EEOICPA.

For DOE employees with cancer, the DOE facility definition only determines eligibility for a dose 

reconstruction, which is a prerequisite to a compensation decision (except for members of the Special 

Exposure Cohort). The compensation decision for cancer claimants is based on a section of the 

statute entitled “Exposure in the Performance of Duty.” That provision (42 U.S.C. § 7384n(b)) says 

that an individual with cancer “shall be determined to have sustained that cancer in the performance 

of duty for purposes of the compensation program if, and only if, the cancer ... was at least as likely 

as not related to employment at the facility [where the employee worked], as determined in 

accordance with the [probability of causation] guidelines established under subsection (c) ....” 42 

U.S.C. § 7384n(b). Neither the statute nor the probability of causation guidelines (nor the dose 

reconstruction regulation) define “performance of duty” for DOE employees with a covered cancer or 

restrict the “duty” to nuclear weapons work.

As noted above, the statute includes a definition of a DOE facility that excludes “buildings, structures, 

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

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

While this definition contains an exclusion with respect to the Naval Nuclear Propulsion Program, the 

section of EEOICPA that deals with the compensation decision for covered employees with cancer 

(i.e., 42 U.S.C. § 7384n(b), entitled “Exposure in the Performance of Duty”) does not contain such an 

exclusion. Therefore, the statute requires NIOSH to include all radiation exposures in its dose 

reconstructions for employees at DOE facilities, including radiation exposures related to the Naval 

Nuclear Propulsion Program. As a result, all internal and external dosimetry results are considered 

valid for use in dose reconstruction. No efforts are made to determine the eligibility of any fraction of 

total measured exposure for inclusion in dose reconstruction.

This Site Profile documents historical practices at the Lawrence Livermore National Laboratory (LLNL) 

and provides information for the evaluation of internal and external dosimetry data for unmonitored 

and monitored workers; it can serve as a supplement to, or substitute for, individual monitoring data.

Occupational environmental dose refers to exposures received by workers outside the facilities at the 

Lawrence Livermore National Laboratory (LLNL) from elevated ambient radiation, from facility 

discharges to the environment, and from resuspension of radionuclides in soils. Effluents can result in 
Document No. ORAUT-TKBS-0035-4 Revision No. 00 Effective Date: 11/18/2005 Page 7 of 27

internal and external exposures by inhalation of airborne radionuclides, ingestion of radionuclides, and 

exposure to external radiation. This TBD describes the estimated annual intakes for inhalation 

exposure and estimated radiation dose as a result of ambient exposures at LLNL. Environmental 

measurements do not distinguish the source of emissions and, therefore, will reflect air concentrations 

from nearby as well as distant sources. The estimates of emissions were useful in filling in some 

gaps in measurement data and were critical to estimating exposures before the start of 

comprehensive and routine measurement data reports.

4.1 INTRODUCTION

LLNL began operations in 1952, when it was known as the University of California Radiation 

Laboratory, a branch of what is now the Lawrence Berkeley National Laboratory, or Lawrence 

Radiation Laboratory (LRL) at Livermore, California. Throughout its history, LLNL has processed and 

handled a variety of radionuclides, including uranium and transuranic elements, mixed fission 

products, and accelerator-produced isotopes. This TBD discusses two sites: the main site, which 

conducts research and development activities, and Site 300, where explosive testing experiments 

occur. Site 300 began operations in 1955. There was a potential for internal dose from the breathing 

of airborne concentrations of radionuclides released on the LLNL site; Section 4.2 of this TBD 

contains detailed information about the collection and analysis of air samples and how this information 

is used for dose reconstruction. There was a potential for ingestion of radioactive materials from 

resuspension of radioactive materials (Section 4.3) and the ingestion of water potentially containing 

tritium (Section 4.4). Section 4.5 describes the potential external dose from sources of radiation 

outside and inside the process buildings and Section 4.6 discusses possible sources of uncertainty 

associated with each of these sources of environmental dose.

The receptors of concern addressed in this TBD were unmonitored workers, namely LLNL employees 

who did not wear external dosimetry or who were not monitored for internal exposures. At LLNL, 

everyone, including administrative personnel, wore some form of external radiation dosimetry as of 

March 1958 (Nolan 1958). To provide the basis for estimating the environmental dose for years when 

monitoring did not occur or was not sufficient to apply to coworkers, this TBD provides annual intakes 

and ambient external dose from 1952 (1955 for Site 300) to 2001 (the last year with publicly available 

data). 

4.2 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE CONCENTRATION

4.2.1

4.2.1.1 Pre 1971

Ambient Air Sample Collection Network

Air samples were collected in 1961 in eleven locations around the main site, including two perimeter 

locations (LRL, 1962). The two perimeter locations, southeast perimeter and west perimeter, were 

collected continuously for a seven day period on HV-70 paper, measuring 4x9 inches. The other air 

sampling stations were located at a variety of distances ranging from the perimeter to as far as five 

miles from the perimeter. As shown on Table 4-1, sample locations SALV and MESQ were 

established as early as 1961, for the purposes of perimeter air monitoring. At Site 300, six air 

sampling locations were established in 1962; each of the locations were located within the boundary 

of the Site 300 perimeter. The air sampling equipment located at Site 300 was equivalent to the 

equipment used at the main site. As shown in Table 4-2, sample locations Bunker 801E, ECP, 

EOBS, GOLF, WCP and WOBS were established in 1962.
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4.2.1.2 Post 1971

In 1971, LLNL established a network of permanent stations to collect air samples from the perimeter 

of the facility and other locations inside the facility (Gallegos and Surano 1992). LLNL collected and 

analyzed air samples for gross alpha and beta radiations, 3

H, 239Pu and 240Pu, and 235U and 238U. The 

existing monitoring networks were established in 1992 for surveillance of air particulates and tritium in 

the environs of the Livermore site and Site 300, as well as in the surrounding Livermore Valley and 

the City of Tracy (Gallegos and Surano 1992). Tables 4-1 and 4-2 list the sampling locations for the 

monitoring network at LLNL and Site 300, respectively, and Figures 4-1 and 4-2 show these locations. 

Table 4-1. Ambient air sampling station locations within LLNL boundary. 

Designation Location

Direction from 

center of LLNL

COW Located at security fence north of outer loop road N

MET Meteorological station NW

VIS At security fence near east entrance E

CAFÉ At security fence near south entrance S

SALV At security fence near southeast perimeter fence SE

MESQ At security fence near west entrance W

CRED Solar evaporators for plutonium waste processing; 

representative of maximum exposed individual

SE

Table 4-2. Ambient air sampling station locations at Site 300. 

Designation Location

Direction from 

center of Site 300

COHO Near south perimeter of facility; representative of 

maximum exposed individual

S

EOBS Northeast of Bunker 801E NE

ECP Center of Site 300 E

WCP Test Area W

GOLF South perimeter near fence S

TFIR Offsite location in City of Tracy E

NPS North of Bunker 801E N

WOBS West of Bunker 850 W

801E East of Bunker 801E E

4.2.1.3 Sampling Locations

The principal purpose of the ambient air monitoring network was to assess if air emissions from LLNL 

affected the air quality in the surrounding area. Air samples were collected from locations where a 

significant concentration of effluents from LLNL operations could be reasonably detected regardless 

of local meteorology (LLNL 1986). This demonstrated compliance with U.S. Department of Energy 

(DOE) Derived Concentration Guidelines or U.S. Environmental Protection Agency (EPA), State of 

California, and (as of 1997) U.S. Nuclear Regulatory Commission regulations for airborne releases to 

the general public around LLNL. There were seven onsite monitoring stations.1

 

1 An air sampling station was added on the east side of LLNL in 2000. This station, CRED, was near the area representative 

of the maximally exposed individual for LLNL. 

 This TBD considers 

only those ambient air monitoring locations inside the LLNL security fence for worker intake. 

Table 4-1 lists these monitors and their locations.
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Figure 4-1. Air particulate and tritium sampling locations on Livermore site

(Source: LLNL 2002).

Air samplers were positioned to provide reasonable probability that any significant concentration of 

radioactive effluents from LLNL operations was detected. The Livermore site radiological air 

particulate sampling network consisted of seven samplers at the perimeter with one (CRED) serving 

as the site-wide maximally exposed individual (SW-MEI) as reported for National Emissions 

Standards for Hazardous Air Pollutants (NESHAPS) monitoring. CRED was in the southeast 

quadrant in an area of known plutonium contamination attributed to historical operations, which 

included the operation of solar evaporators for plutonium-containing liquid waste.
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Figure 4-2. Air particulate and tritium sampling locations at Site 300 and off site (Source, LLNL, 

2002).

The Site 300 air particulate monitoring network included eight sampling units placed around the site 

and near firing tables and one in downtown Tracy. Site 300 is in a remote area and access is limited. 

LLNL based the selection of the monitoring sites on safety, power, and access considerations. The 

COHO location served as the SW-MEI for NESHAPS reporting purposes. LLNL added two sampling 

systems in the Livermore Valley in July 1997 as part of the new low-volume radiological air particulate 

sampling network (LLNL 2001). These samplers were generally upwind of the Livermore site. LLNL 

used the results to establish background levels of gross alpha and beta activity for direct comparison 

to emissions from the air effluent samplers. 

In addition, LLNL maintained 12 continuously operating airborne tritium samplers on the Livermore 

site to assess current activities that influence environmental impacts. These stations were deployed 

in 1973; specific stacks were sampled for tritium as early as 1971 (Gede and Gildea, 1980) (SAIC 

1993). 

4.2.1.4 Radionuclides of Significance

The LLNL environmental monitoring program identified radionuclides of significance. The program 

analyzed air samples for the presence of gross alpha and gross beta radiations, and for specific 

isotopes that represented more than 90% of the LLNL radioactive materials inventory. Specifically, it 

analyzed air samples for the presence of tritium, 239Pu, and isotopes of uranium. Tables 4-3 (a and b) 
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and 4-4 (a and b) list the results for 234U because the radiation dose conversion factor was greater for 

that isotope than for 238U and 235U. For the periods of 1961 through 1971, the gross alpha 

concentration should be used to calculate the presence of 239Pu and a second calculation using 234U; 

the assigned dose should be selected from the higher of the two calculations. The gross beta 

concentration should be assumed to be thorium 234.

For Site 300, radionuclides of significance were selected according to the inventory of radioactive 

materials and the type of work performed at that site. Specifically, the presence of depleted uranium 

(DU) indicated the analysis of 238U and 235U. For the purpose of this TBD, the dose reconstructor 

should assign the dose associated with 235U and 234U for depleted uranium. For the periods of 1961 

through 1971, the gross alpha concentration for Site 300 should be used to calculate the presence of 

235U and a second calculation using 234U; the assigned dose should be selected from the higher of the 

two calculations. The gross beta concentration should be assumed to be thorium 234.

4.2.2

LLNL used several different networks, each representing a general location and type of analysis to 

perform environmental air sampling. There were separate networks for sampling radiological 

particulates and beryllium particulates at the Livermore site and Site 300 as well as a low-volume 

radiological air sampling network and a tritium sampling network in Livermore and one tritium 

sampling location at Site 300. Four collection media were employed: glass-fiber filters for radiological 

particulates, cellulose filters for beryllium particulates, membrane filters for low-volume radiological 

particulates, and silica gel for tritium. All monitoring networks used continuously operating samplers 

(LLNL 1992a). 

Methodology

As outlined in the Environmental Regulatory Guide for Radiological Effluent Monitoring and 

Environmental Surveillance (DOE 1991), data for gross alpha, gross beta, and gamma isotopes 

detected on air filters were used only as trend indicators; specific radionuclide analysis occurred for 

plutonium and uranium, depending on the location. All analytical results were reported as a measured 

concentration per volume of air. When activity was less than the minimum detection concentration, 

the calculated value was reported. Particle size distributions were not determined because the 

estimated effective dose equivalent to the MEI is well below the 0.01 mSv (1 mrem) allowable limit 

(DOE 1991). The analytical laboratory used 230Th and 90Sr as calibration sources to determine alpha 

and beta counting efficiencies, respectively. Annual counting efficiency measurements were made for 

each detector. Periodic cross-checks used standards certified by the EPA. Background and 

efficiency checks occurred daily, and a matrix and method blank were run with every batch of 20 

samples. LLNL kept records of background and counting efficiency variations that occurred in the 

counting equipment. The analytical laboratory reported the actual instrumentation values, including 

negative results that occurred when background measurements were higher than those for the filters.

Concentrations of various airborne radionuclides were measured at Livermore site perimeters, at 

offsite locations near the Livermore site, and at Site 300. Prior to 1964, air samples were collected 

from two perimeter locations and nine offsite locations on 4-in. by 9-in. HV-70 paper at a sample flow 

rate of approximately 4 ft3

/min of air (LRL, 1964). The samples were collected for a period of 7 days. 

After a 4-day delay for decay of the radon/thoron progeny, gross alpha and beta activities on the filters 

were determined with an automatic gas flow proportional counter. Monthly composites of perimeter 

filters were counted for specific gamma-emitting radionuclides using a Ge(Li) detector equipped with 

Compton suppression. Following gamma counting, the perimeter filters were grouped by sampling 

locations. The individual samples were analyzed for the presence of 239Pu, 238Pu, 137Cs, 235U, and 

238U. In July 1964, four sample locations were added on the perimeter; six locations around the site 
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perimeter were sampled (LRL, 1964). The six samplers on the Laboratory perimeter used 5.2 × 10-2

m2 Whatman-41 cellulose filters. The particulates were collected on Whatman filters using an 

average air flow rate of 700 L/min (Lindeken and Wang, 1978). Tritium was collected with columns 

packed with silica gel, with an air flow rate of 0.5 L/min. Particulate filters were changed each week at 

all locations, and tritium samplers were changed every 2 weeks. From 1973 to 1976, area samples 

for tritium were exchanged each week. After 1976, tritium samples were exchanged every 2 weeks 

(SAIC 1993). Duplicate quality control samplers operated in parallel with the permanent sampler at a 

given site, and samples from these duplicates were analyzed to confirm results (Lindeken and Wang, 

1978). 

In April 1997, the radiological air particulate sampling filter media changed from cellulose to glass 

fiber. However, blank glass-fiber filters contain nontrivial amounts of naturally occurring radiological 

isotopes (Althouse 1998) including 235U, 238U, 40K, 226Ra, 228Ra, and 228Th. In fact, the amount of these 

naturally occurring isotopes in the filters was often greater than the amount of the isotopes being 

filtered from the air. The filters were cut in half, and half of each filter is retained for beryllium 

analysis. 

Gross alpha and gross beta activities were determined by gas flow proportional counting, plutonium 

and uranium isotopes by alpha spectrometry, gamma emitters by gamma spectroscopy, and tritium by 

liquid scintillation. In 2001, a correction factor was applied to tritium concentrations to account for 

dilution of the collected tritium from air moisture by an unknown quantity of water in supposedly dry 

silica gel (LLNL 2001). On average, the corrected concentrations were 1.6 times higher than 

uncorrected concentrations. 

Gross alpha, gross beta, and gamma emitters on air filters were used as trend indicators; specific 

radionuclide analysis was completed for plutonium and uranium after 1971. Radiological analytical 

results were reported as a measured activity per volume of air. Particle size distributions on air 

samples were not determined because the estimated effective dose equivalent to the maximally

exposed individual (from the total particulate) was below 10 mrem committed effective dose 

equivalent (CEDE; LLNL 2001).

Portions of the glass-fiber filters from the Livermore locations were analyzed for the presence of 

239+240Pu. Similarly, portions of the glass-fiber filters from Site 300 were analyzed for the presence of 

235U and 238U. The filters were placed in a muffled furnace to reduce organic content and then 

dissolved in a mixture of nitric and hydrochloric and/or hydrofluoric acids. Plutonium and uranium 

were separated by an ion exchange process. Each separated element was purified further by ion 

exchange. It was then electroplated onto a stainless-steel disk and analyzed by alpha spectrometry. 

For gamma scanning, a site composite was created using all weekly glass-fiber filters for Site 300 

perimeter locations (801E, ECP, EOBS, GOLF, NPS, WCP, and WQBS). This composite was 

prepared for analysis in the same manner as that used for plutonium and uranium samples. After it 

was muffled and digested, it was counted for more than 40 gamma-emitting radionuclides using 

Ge(Li) detectors. In addition to gamma scanning, the Site 300 composite was analyzed for plutonium. 

Duplicate radiological quality assurance (QA) samples were processed to confirm the precision of the 

analytical results obtained from the samplers. A duplicate QA sampler was operated for 2 months in 

parallel with the permanent sampler at a given site. In addition, a trip blank was collected during each 

route. The QA trip blanks and QA duplicates were processed in the same manner as that used for 

routine samples and analyzed for the same radiological parameters. 
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Table 4-3a. Maximum site-wide 

annual median intakes (Bq/yr) via 

inhalation.

a 

Year

Gross Alpha Gross Beta

Conc 

(Bq/m3)

Intake

(Bq/yr)

Conc 

(Bq/m3)

Intake

(Bq/yr)

1961 3.1E-4 0.9 9.3E-3 28

1962 1.7E-4 0.5 9.0E-3 27

1963 1.2E-4 0.4 1.3E-1 390

1964 1.2E-4 0.4 2.4E-2 72

1965 3.7E-5 0.1 6.5E-3 20

1966 3.7E-5 0.1 2.6E-3 7.8

1967 2.9E-5 0.09 4.7E-3 14

1968 3.7E-5 0.1 4.4E-3 13

1969 3.7E-5 0.1 3.8E-3 11

1970 3.7E-5 0.1 5.6E-3 17

1971 7.5E-5 0.2 1.5E-2 45

a. Assume an inhalation rate of 20 L/min for 

2,500 hr per year.

Table 4-3b. Maximum site-wide annual median intakes (Bq/yr) 

via inhalation.

a

Year

Tritium Pu-239+240 U-234

Conc 

(Bq/m3)

Intakeb

(Bq/yr)

Conc 

(Bq/m3)

Intake 

(Bq/yr)

Conc 

(Bq/m3)

Intake 

(Bq/yr)

1972 1.4c 6.4E3 1.3E-6c 3.9E-3 1.3E-6c 3.9E-3

1973 1.4 6.4E3 1.3E-6c 3.9E-3 1.3E-6c 3.9E-3

1974 1.7 7.8E3 1.3E-6c 3.9E-3 1.3E-6c 3.9E-3 

1975 1.7 7.8E3 1.3E-6c 3.9E-3 1.3E-6c 3.9E-3

1976 3.0 1.4E4 1.0E-6 3.0E-3 1.1E-6 3.3E-3

1977 2.2 1.0E4 1.3E-6c 3.9E-3 1.1E-6c 3.3E-3 

1978 2.0 9.0E3 1.3E-6c 3.9E-3 1.1E-6c 3.3E-3

1979 1.6 7.1E3 8.5E-7 2.5E-3 1.1E-6 3.3E-3

1980 1.4 6.4E3 5.2E-7 1.5E-3 9.2E-7 2.8E-3

1981 1.6 7.1E3 5.0E-7c 1.5E-3 1.1E-6c 3.3E-3

1982 1.4 6.4E3 5.0E-7c 1.5E-3 1.1E-6c 3.3E-3

1983 1.1 4.9E3 5.0E-7c 1.5E-3 1.1E-6c 3.3E-3

1984 1.3 5.9E3 5.0E-7c 1.5E-3 1.1E-6c 3.3E-3 

1985 1.0 4.5E3 2.2E-7 6.6E-4 1.1E-6 3.3E-3

1986 0.9 4.1E3 1.5E-7 5.6E-4 1.1E-6 3.3E-3

1987 1.7 7.8E3 1.9E-7 5.8E-4 1.3E-6 3.9E-3 

1988 1.3 5.9E3 3.3E-8 9.9E-5 2.7E-6 8.1E-3

1989 0.9 4.1E3 7.0E-8c 2.1E-4 2.7E-6c 8.1E-3

1990 0.5 2.3E3 7.0E-8c 2.1E-4 2.7E-6c 8.1E-3

1991 0.7 3.3E3 7.0E-8 2.1E-4 1.9E-6 5.8E-3

1992 0.2 9.0E2 6.8E-8 2.0E-4 1.1E-6 3.3E-3

1993 0.2 9.0E2 2.7E-8 8.1E-5 8.7E-7 2.6E-3

1994 0.1 4.5E2 3.4E-8 1.0E-4 6.8E-7 2.0E-3 

1995 0.07 3.3E2 3.5E-8 1.0E-4 6.9E-7 2.1E-3

1996 0.2 9.0E2 2.4E-8 7.3E-5 7.7E-7 2.4E-3

1997 0.1 4.5E2 8.8E-9 2.6E-5 6.6E-7 2.0E-3 

1998 0.09 4.1E2 8.6E-9 2.6E-5 4.6E-7 1.4E-3

1999 0.09 4.1E2 5.5E-9 1.6E-5 8.9E-7 2.6E-3

2000 0.05 2.3E2 9.1E-9 2.8E-5 1.3E-8 3.9E-5

2001 0.04 1.8E2 9.1E-9 2.8E-5 1.3E-8 3.9E-5

a. Assume an inhalation rate of 20 L/min for 2,500 hr per year.

b. Assume total intake is the sum of inhalation and skin absorption. Total 

intake is 1.5 × the inhalation intake.

c. The actual concentration for this isotope was not reported. The result is a 

maximum value reported for a different year. See Section 4.2.3 for 

additional information.
Document No. ORAUT-TKBS-0035-4 Revision No. 00 Effective Date: 11/18/2005 Page 14 of 27

Table 4-4a. Site 300 maximum annual 

median intakes (Bq/yr) via inhalation.

a 

Gross Alpha Gross Beta

Conc 

(Bq/m3)

Intake

(Bq/yr)

Conc 

(Bq/m3)

Intake

(Bq/yr)

1961 4.8E-4 1.4 2.6E-3 7.8

1962 2.1E-4 0.6 1.4E-1 420

1963 1.0E-4 0.3 1.6E-1 480

1964 1.3E-4 0.4 2.2E-2 66

1965 2.6E-5 0.1 6.3E-3 19

1966 1.9E-5 0.06 3.2E-3 9.6

1967 1.5E-5 0.05 5.6E-3 17

1968 7.4E-5 0.2 5.6E-3 17

1969 7.4E-5 0.2 5.7E-3 17

1970 7.4E-6 0.02 8.6E-3 26

1971 3.7E-5 0.1 1.7E-2 51

a. Assume an inhalation rate of 20 L/min for 

2,500 hr per year.

Table 4-4b. Site 300 maximum annual median intakes (Bq/yr) via 

inhalation.

a

Year

U-238 U-235 Pu-239+240

Conc 

(Bq/m3)

Intake

(Bq/yr)

Conc 

(Bq/m3)

Intake 

(Bq/yr)

Conc 

(Bq/m3)

Intake 

(Bq/yr)

1972 2.0E-6b 6.0E-3 4.9E-8b 1.5E-4 6.2E-6b 1.9E-2

1973 2.0E-6b 6.0E-3 4.9E-8b 1.5E-4 6.2E-6b 1.9E-2

1974 2.0E-6b 6.0E-3 4.9E-8b 1.5E-4 6.2E-6b 1.9E-2

1975 2.0E-6b 6.0E-3 4.9E-8b 1.5E-4 6.2E-6b 1.9E-2

1976 2.0E-6 6.0E-3 4.8E-8 1.4E-4 6.2E-6 1.9E-2

1977 2.0E-6b 6.0E-3 4.9E-8b 1.5E-4 6.2E-6b 1.9E-2

1978 2.0E-6b 6.0E-3 4.9E-8b 1.5E-4 6.2E-6b 1.9E-2

1979 1.5E-6 4.5E-3 4.9E-8 1.5E-4 6.2E-6b 1.9E-2

1980 1.8E-6 5.4E-3 9.3E-8 2.8E-4 8.5E-7 2.6E-3

1981 1.8E-6b 5.4E-3 9.3E-8b 2.8E-4 8.5E-7b 2.6E-3

1982 1.8E-6b 5.4E-3 9.3E-8b 2.8E-4 8.5E-7b 2.6E-3

1983 1.8E-6b 5.4E-3 9.3E-8b 2.8E-4 8.5E-7b 2.6E-3

1984 1.8E-6b 5.4E-3 9.3E-8b 2.8E-4 8.5E-7b 2.6E-3

1985 1.0E-6 3.0E-3 3.8E-8 1.1E-4 3.0E-7 3.0E-4

1986 8.3E-7 2.5E-3 3.8E-8 1.1E-4 3.9E-8 1.2E-4

1987 2.8E-6 8.4E-3 4.7E-8 1.4E-4 5.9E-8 1.8E-4

1988 2.7E-6 8.1E-3 6.2E-8 1.9E-4 3.9E-7 1.2E-3

1989 3.4E-6b 1.0E-2 7.3E-8b 2.2E-4 3.9E-7b 1.2E-3

1990 3.4E-6b 1.0E-2 7.3E-8b 2.2E-4 3.9E-7b 1.2E-3

1991 3.4E-6 1.0E-2 7.3E-8 2.2E-4 7.8E-8 2.3E-4

1992 3.0E-6 9.0E-3 1.4E-7 4.2E-4 1.4E-8 4.2E-5

1993 1.9E-6 5.7E-3 6.1E-9 1.8E-5 7.3E-9 2.2E-5

1994 1.9E-6 5.7E-3 6.1E-8 1.8E-4 7.3E-9 2.2E-5

1995 1.7E-7 5.1E-4 8.2E-8 2.5E-4 1.2E-8 3.6E-5

1996 2.4E-6 7.2E-3 3.9E-8 1.1E-4 1.0E-8 3.0E-5

1997 2.2E-6 6.6E-3 1.0E-7 3.0E-4 1.7E-8 5.1E-5

1998 7.0E-6 2.1E-2 1.2E-7 3.6E-4 1.6E-8 4.8E-5

1999 2.2E-6 6.6E-3 7.3E-8 2.2E-4 8.7E-9 2.6E-5

2000 3.9E-6 1.7E-2 9.4E-8 2.8E-4 1.9E-8 5.7E-5

2001 8.8E-6 2.6E-2 4.6E-7 1.4E-3 9.8E-9 2.9E-5

a. Assume an inhalation rate of 20 L/min for 2,500 hr per year.

b. The actual concentration for this isotope was not reported. The result is a maximum value 

reported for a different year. See Section 4.2.3 for additional information.
Document No. ORAUT-TKBS-0035-4 Revision No. 00 Effective Date: 11/18/2005 Page 15 of 27

4.2.3

This section discusses the air monitoring results from all air surveillance locations at the Livermore 

site and Site 300. This TBD relies on several data sources for estimating ambient airborne 

radionuclide concentrations and for annual airborne releases by radionuclide. The principal sources 

were the annual environmental reports for LLNL from 1976 and through 2001 (see the reference list at 

the end of this document). The analysis for this TBD collected measurements from the air monitoring 

locations listed in Tables 4-1 and 4-2 from these documents and applied them to the methodology 

described above. If data were not available or additional information was required, these documents 

were supplemented by information provided by DOE or the University of California (the DOE 

contractor at LLNL). 

Estimation of Annual Intake from Airborne Radionuclides

Factors other than transport and release rates influenced the gradual reduction in the air 

concentration over the history of LLNL. Over the 40 years of monitoring, concentrations of airborne 

radioactive materials decreased by more than 3 orders of magnitude. These factors include: 

• Improved monitoring methods reducing the minimum detectable concentrations

• Naturally occurring radioactive material released from a nearby fossil generating plant that has 

reduced releases over the years through improved environmental controls

• The effect of atmospheric weapons testing and the radioactive decay of fallout

The methodology for the intake from onsite atmospheric radionuclide concentrations applied directly

to available air sampling data. Air sampling measurement error and uncertainty were evaluated by 

using the maximum value reported for the year at any location. Available data were limited to annual 

averages for each location; statistical analysis was not possible due to the limited amount of data.

The following factors limited estimates of airborne concentrations at specific locations around the 

LLNL site using traditional transport modeling approaches:

• The numerous release points, which include stacks, vents, and other emission sources;

• The characteristics of the release points;

• The limited number of air sampling locations;

• The relatively short distances between the release points and the onsite receptor locations; 

and

• The density and configurations of buildings at the site.

4.2.3.1 Potential Intakes Prior to 1961

Prior to 1961, air monitoring data were not available for review. Estimated worker intakes in this 

revision of the TBD are restricted to post 1960 years, beginning in 1961. Efforts are currently in 

progress for developing intakes that occurred prior to 1961 and may be presented in a future revision 

of this TBD.
Document No. ORAUT-TKBS-0035-4 Revision No. 00 Effective Date: 11/18/2005 Page 16 of 27

4.2.3.2 Potential Intakes 1961 to 1971

Data from Environmental Reports were available for the years 1961 through 1971 (LRL 1961, 1963, 

1964, 1965, 1966, 1967, 1968, 1969, 1970, 1971). The air monitoring data provided no isotopic 

analysis; the data were reported only for the gross alpha and gross beta radiations. These data are 

provided in Table 4.3a and 4.4a for the LLNL site and Site 300, respectively. As described in Section 

4.2.1.2, the isotopes of significance at LLNL were 239+240Pu and 234U, both decaying by alpha radiation. 

Isotopes decaying by beta radiation were determined to be progeny of these two parent isotopes. 

Tritium environmental air monitoring data were not available for review. Estimated worker intakes of 

tritium in this revision of the TBD are restricted to the years beginning with 1972. Efforts are currently 

in progress for developing tritium intakes that occurred prior to 1972 and may be presented in a future 

revision of this TBD.

The isotopes of significance at Site 300 are 238U and 235U, both decaying by alpha radiation. The 

presence of beta radiation indicates the progeny of uranium. This TBD assumed that a worker 

inhaled 20 L of air per minute for 2,500 hr per year or approximately 3 × 106 L of air per year (3,000 

m3 of air per year). The air monitoring data indicated that potential inhalation exposures at the LLNL 

site to gross alpha radiation ranged from 3x10-5 to 3x10-4 Bq/m3 or a potential inhalation exposure 

from 1 to 9 Bq/year. The air monitoring data indicated that potential inhalation exposures at Site 300 

to gross alpha radiation ranged from 3x10-5 to 3x10-4 Bq/m3 or a potential inhalation exposure from 1 

to 9 Bq/year.

4.2.3.3 Potential Intakes post 1971

Past estimates of offsite doses to members of the public indicated that the potential internal dose from 

airborne releases to LLNL workers should be relatively low (10 mrem CEDE or less) (LLNL 1977, 

1980, 1987, 1988, 1989, 1992a,b, 1993, 1994, 1995, 1996, 1997, 1998, 1999, 2000, 2001, 2002). Air 

data from nearby locations and at the security fence were consistent from location to location, but 

dependent on year. This provided reasonable approximation of general airborne radioactivity and 

establishment of trends as a function of time. 

Samples were collected and analyzed for airborne radionuclides at the perimeter of the facility. 

Table 4-3a lists concentrations for 3

H, 239Pu, and 234U. This TBD assumed that a worker inhaled 20 L 

of air per minute for 2,500 hr per year or approximately 3 × 106 L of air per year (3,000 m3 of air per 

year). The concentration of tritium ranged from 0.04 to 3 Bq/m3

; the potential for exposure to tritium 

ranged from 180 to 14,000 Bq/yr. For tritium, the total uptake by the body increased by 50% to 

account for absorption by the skin in addition to the inhalation pathway (Gede and Gildea, 1980). The 

concentration of 239Pu ranged from 9.1 × 10-9 to 1 × 10-6 Bq/m3

; the potential for exposures to 239Pu 

ranged from 0.00003 to 0.003 Bq/yr. Figure 4-3 shows how the concentration of 239+240Pu decreased 

from 1982 through 2001. The concentration dropped by more than 20 times over the 20 years of 

sampling. The concentration of 234U ranged from 1.3 × 10-8 to 1× 10-6 Bq/m3

; the potential for 

exposures to 234U ranged from 0.00004 to 0.003 Bq/yr.

Tables 4-4a and b indicate the potential for exposure to inhalation of airborne radionuclides at Site 

300. The concentration of 238U ranged from 8.3 × 10-7 to 3 × 10-6 Bq/m3

; the potential for exposure to 

238U ranged from 0.009 to 0.01 Bq/yr. The concentration of 235U ranged from 6 × 10-9 to 1.0 × 10-7

Bq/m3

; the potential for exposure to 235U ranged from 1.8 × 10-5 to 0.001 Bq/yr. The concentration of 

239Pu ranged from 9.8 × 10-9 to 1 × 10-6 Bq/m3

; the potential for exposures to 239Pu ranged from 2.6 × 

10-5 to 0.003 Bq/yr. 
Document No. ORAUT-TKBS-0035-4 Revision No. 00 Effective Date: 11/18/2005 Page 17 of 27

Figure 4-3. Calculated annual median concentrations of 239+240Pu, 1982-2001 (Source: LLNL 2002).

In years for which air concentration data were unavailable, release data and maximum air 

concentration data were evaluated and estimates of the air concentration were generated. Tables 4-3 

(a and b) and 4-4 (a and b) list these data, and demonstrate that chosen values were reasonable and 

claimant-favorable. Using the maximum airborne radionuclide concentration in a year (either from 

actual data or an estimated value), claimant-favorable annual intakes for the radionuclides of concern 

were derived by using an assumed individual’s annual respiration rate of 3,000 m3

/yr. Most years 

have only a potential, even by being claimant-favorable, for contributing approximately 10 mrem 

(CEDE) or less.

4.3 INTAKE FROM DRINKING WATER

The annual environmental reports discuss concentrations of radionuclides in the drinking water at 

LLNL (LRL 1966, 1967, 1968, 1969, 1970, 1970, 1971) (LLNL 1977, 1980, 1987, 1988, 1989, 

1992a,b, 1993, 1994, 1995, 1996, 1997, 1998, 1999, 2000, 2001, 2002). Several water sources were 

sampled and analyzed; BELL, GAS, PALM, ORCH, and TAP were identified as sources of drinking 

water. The water from these locations was sampled at least once per year. The POOL sample 

location was the onsite swimming pool. The median activity in the drinking water was estimated from 

calculated values and found to be below the minimum analytical detectable activity. Concentrations of 

tritium in the POOL location ranged from 0.8 to 200 Bq/L; the maximum concentration was reported in 

1988. The LLNL swimming pool is close to the main sources of tritium at LLNL (LLNL 2002). This 

TBD assumed that a worker at LLNL ingested 3 L of water per day or approximately 1,100 L/yr. As a 

consequence, the estimated potential for ingestion of tritium ranged from 880 to 220,000 Bq/yr. Table 

4-5 lists the potential for ingestion of tritium over time. The values in the tables before 1966 were 

extrapolated and intended to be claimant favorable.
Document No. ORAUT-TKBS-0035-4 Revision No. 00 Effective Date: 11/18/2005 Page 18 of 27

Table 4-5. Maximum site-wide annual intakes (Bq/yr) via drinking water. 

Year

Tritium

Conc 

(Bq/l)

Intakea

(Bq/yr)

1952 100b 1.1E5

1953 100b 1.1E5

1954 100b 1.1E5

1955 100b 1.1E5

1956 100b 1.1E5

1957 100b 1.1E5

1958 100b 1.1E5

1959 100b 1.1E5

1960 100b 1.1E5

1961 100b 1.1E5

1962 100b 1.1E5

1963 100b 1.1E5

1964 100b 1.1E5

1965 100b 1.1E5

1966 185 2.0E5

1967 185 2.0E5

1968 185 2.0E5

Year

Tritium

Conc 

(Bq/l)

Intakea

(Bq/yr)

1969 185 2.0E5

1970 200 2.2E5

1971 100b 1.1E5

1972 100b 1.1E5

1973 100b 1.1E5

1974 100b 1.1E5

1975 100b 1.1E5

1976 26 2.9E4

1977 26b 2.9E4

1978 26b 2.9E4

1979 11 1.2E3

1980 26b 2.9E4

1981 26b 2.9E4

1982 26b 2.9E4

1983 26b 2.9E4

1984 26b 2.9E4

1985 7.4 8.1E3

Year

Tritium

Conc 

(Bq/l)

Intakea

(Bq/yr)

1986 7.4 8.1E3

1987 3.7 4.1E3

1988 6.6 7.2E3

1989 7b 7.7E3

1990 7b 7.7E3

1991 4.5 5.0E3

1992 4.4 4.8E3

1993 1.0 1.1E3

1994 2.8 3.1E3

1995 0.8 8.8E2

1996 2.1 2.3E3

1997 2.4 2.6E3

1998 1.2 1.3E3

1999 2.2 2.4E3

2000 2.0 2.2E3

2001 3.0 3.3E3

a. Assume intake of 3 L of water per day; 1,100 L ingested per year.

b. Result was extrapolated from available annual environmental reports from other years.

No data were available for drinking water at Site 300; this TBD assumed that the concentration of 

tritium was the same as the source of drinking water at LLNL. Table 4-6 lists estimates for Site 300; 

the data were identical to the concentrations listed in Table 4-5.

Table 4-6. Site 300 maximum annual intakes (Bq/yr) via drinking water.

a 

Year

Conc 

(Bq/l)

Intakeb

(Bq/yr)

1955 100b 1.1E5

1956 100b 1.1E5

1957 100b 1.1E5

1958 100b 1.1E5

1959 100b 1.1E5

1960 100b 1.1E5

1961 100b 1.1E5

1962 100b 1.1E5

1963 100b 1.1E5

1964 100b 1.1E5

1965 100b 1.1E5

1966 185 2.0E5

1967 185 2.0E5

1968 185 2.0E5

1969 185 2.0E5

1970 200 2.2E5

Year

Conc 

(Bq/l)

Intakeb

(Bq/yr)

1971 100b 1.1E5

1972 100b 1.1E5

1973 100b 1.1E5

1974 100b 1.1E5

1975 100b 1.1E5

1976 26 2.9E4

1977 26b 2.9E4

1978 26b 2.9E4

1979 11 1.2E3

1980 26b 2.9E4

1981 26b 2.9E4

1982 26b 2.9E4

1983 26b 2.9E4

1984 26b 2.9E4

1985 7.4 8.1E3

1986 7.4 8.1E3

Year

Conc 

(Bq/l)

Intakeb

(Bq/yr)

1987 3.7 4.1E3

1988 6.6 7.2E3

1989 7b 7.7E3

1990 7b 7.7E3

1991 4.5 5.0E3

1992 4.4 4.8E3

1993 1.0 1.1E3

1994 2.8 3.1E3

1995 0.8 8.8E2

1996 2.1 2.3E3

1997 2.4 2.6E3

1998 1.2 1.3E3

1999 2.2 2.4E3

2000 2.0 2.2E3

2001 3.0 3.3E3

a. These data were provided from LLNL site data. No data regarding radioactivity in drinking water were available 

specifically for Site 300.

b. Assume intake of 3 liters of water per day; 1,100 liters ingested per year.
Document No. ORAUT-TKBS-0035-4 Revision No. 00 Effective Date: 11/18/2005 Page 19 of 27

4.4 AMBIENT RADIATION

From 1976 to the present, thermoluminescent dosimeters (TLDs) were used in determining ambient 

radiation levels, including natural background (terrestrial and cosmic) radiation. TLDs were deployed 

at Site 300 in July 1988. Figures 4-4 and 4-5 show the locations of the monitoring stations at the 

Livermore site perimeter and Site 300, respectively. The data in Tables 4-7 and 4-8 were 

summarized from 1989 to 2001 in the annual environmental reports (see the reference list). 

Observations reported in earlier environmental reports, based on surveys with portable instruments, 

state “0.02 mrem/h at all locations.” The instrumentation might have been primitive by today’s 

standards and did not provide continuous monitoring, so these results were overshadowed by recent 

more reliable observations. The results were reported in the annual environmental reports as 

cumulative exposures in millirem per year based on 8,760 hr; Tables 4-7 and 4-8 reflect the data 

provided by LLNL. 

Over the years various combinations of TLD chip configurations and monitoring locations were used. 

From 1988 to the present, minimal changes in external radiation levels were observed at the 

perimeter, as shown in Figure 4-6. Assuming a deep dose equivalent rate of 100 mrem/yr for all 

years would be claimant-favorable and reasonable, and deficiencies in earlier measurement 

techniques would become immaterial.

The analysis for this TBD developed the environmental radiological profile for LLNL for use by dose 

reconstructors when personal dosimetry or bioassay program participation was not required. Site 

annual environmental reports, health physics surveys, and other reports were reviewed for data that 

would be useful in reconstructing ambient radiation levels. 

The ambient radiation measured by TLDs near the security fence included natural background 

radiation, nuclear weapons testing fallout, and cosmic radiation. The TLDs provided an indication of 

worker exposure levels in the general proximity of the security fence but not inside buildings. LLNL 

compared these data annually with TLD data from offsite locations and literature values for State and 

regional exposure levels. The determination has always been that onsite ambient radiological 

conditions as measured at the security fence were not significantly different from offsite, State, and 

regional annual exposure levels. This was attributed to the geology of the region around LLNL. 

4.5 UNCERTAINTY

As discussed in the previous sections, estimates of annual intakes employed conservative (i.e., 

claimant-favorable) assumptions when information was unavailable. For example, applying maximum 

annual exposures to all years was a claimant-favorable assumption, but not overly conservative. 

There was a lack of particle size and solubility. As a consequence, this TBD assumed that all 

effluents were in the respirable size range and selected solubility based on the highest radiation dose 

rate. Effluent data collected in 1976 and later have an estimated uncertainty of ±20%. Data from 

earlier years were less accurate because of less reliable measurement technology and less stringent 

reporting requirements. 

The location of the monitoring points from which data were summarized in this section adds 

uncertainty to the results. The monitoring points, as stated above, have been located around the 

LLNL perimeter and off the site to monitor public exposures. Before 1998, external environmental 

exposures were not monitored in relation to workers; that is, monitoring stations were not normally at 

the interior of the site among the process buildings. Because of data availability, therefore, public 

exposure information had to be used for worker environmental exposures. The maximum value of 
Document No. ORAUT-TKBS-0035-4 Revision No. 00 Effective Date: 11/18/2005 Page 20 of 27

Figure 4-4. Gamma dosimeter locations, Livermore Site, 2001 (Source: LLNL 2002)

environmental exposure was recommended for years when data were unavailable to compensate for 

lack of worker-specific environmental dose information.

The estimated accuracy of the area monitors for external radiation is ±20%. However, subtracting 

background from these measurements adds more uncertainty because of the variability and selection 

of background information. All external environmental dose data were adjusted to reflect a 2,000-hr 

work-year. The data were originally reported in site environmental reports as representative of an 

employee who worked at the site 24 hr a day, 365 days a year. Using a permanently located 
Document No. ORAUT-TKBS-0035-4 Revision No. 00 Effective Date: 11/18/2005 Page 21 of 27

Figure 4-5. Gamma dosimeter locations, Site 300 and vicinity, 2001 (Source: 

LLNL 2002).

employee at the site, however, was an unrealistic assumption that would overstate onsite 

environmental exposures.

Uncertainty related to internal exposures presents similar concerns. The highest internal exposures 

occurred during the early years, and lessened over time with increased controls and better equipment. 

Assumption of the maximum uptake for all years reduces the need to include an uncertainty factor for 

intakes.
Document No. ORAUT-TKBS-0035-4 Revision No. 00 Effective Date: 11/18/2005 Page 22 of 27

Table 4-7. External gamma radiation (mrem/yr). 

Year

Site 

average Error Maximum

1952 88a 10a 100a

1953 88a 10a 100a

1954 88a 10a 100a

1955 88a 10a 100a

1956 88a 10a 100a

1957 88a 10a 100a

1958 88a 10a 100a

1959 50 10 100a

1960 88a 10a 100a

1961 88 10 100a

1962 44 10 100a

1963 50 10 100a

1964 50 10 100a

1965 50 10 100a

1966 88a 10a 100a

1967 20 5 100a

1968 20 5 100a

1969 20 5 100a

1970 20 10 100a

1971 52 5 90

1972 88a 10a 100a

1973 88a 10a 100a

1974 88a 10a 100a

1975 88a 10a 100a

1976 84 2 293

Year

Site 

average Error Maximum

1977 88a 10a 100a

1978 88a 10a 100a

1979 61 3 96

1980 88a 10a 100a

1981 88a 10a 100a

1982 88a 10a 100a

1983 88a 10a 100a

1984 88a 10a 100a

1985 59 6 80

1986 61 5 69

1987 65 7 71

1988 62 7 69

1989 88a 10a 100a

1990 88a 10a 100a

1991 65 6 78

1992 66 5 76

1993 65 5 74

1994 72 9 79

1995 56 8 67

1996 55 4 60

1997 60 1 64

1998 60 2 66

1999 58 3 63

2000 57 2 65

2001 56 1 100

a. External dose for these years was not available. The maximum external dose was reported in 1961. 

Accordingly, the 1961 dose values are used for the years noted.
Document No. ORAUT-TKBS-0035-4 Revision No. 00 Effective Date: 11/18/2005 Page 23 of 27

Table 4-8. Site 300 external gamma radiation (mrem/yr). 

Year

Site 

averagea,b,c Error c Maximum c

1955 100

1956 100

1957 100

1958 100

1959 100

1960 100

1961 100

1962 100

1963 100

1964 100

1965 100

1966 100

1967 100

1968 100

1969 100

1970 100

1971 100

1972 100

1973 100

1974 100

1975 100

1976 100

1977 100

1978 100

Year

Site 

averagea,b,c Error c Maximum c

1979 100

1980 100

1981 100

1982 100

1983 100

1984 100

1985 100

1986 100

1987 100

1988 100

1989 78 7a 89a

1990 78 7a 89a

1991 78 7 89

1992 77 5 87

1993 76 6 85

1994 64 1 74

1995 64 1 74

1996 66 2 72

1997 72 4 77

1998 72 4 78

1999 71 1 80

2000 64 1 70

2001 63 1 66

a. External dose for these years was not available. The maximum external dose was reported in 1991. 

Accordingly, the 1991 dose values are used for the years noted.

b. Environmental radiation monitoring at Site 300 began on July 1, 1988. 

c. Site average (mrem/yr) is based on 10% above maximum value from later years where monitoring was 

performed. Therefore, no maximum or error is provided.

Figure 4-6. Quarterly mean gamma dose measurements at the Livermore site 

perimeter, 1988-2001 (Source: LLNL 2002)
Document No. ORAUT-TKBS-0035-4 Revision No. 00 Effective Date: 11/18/2005 Page 24 of 27

REFERENCES

Althouse, P.E., 1998, Radiological Background Levels found in Glass Fiber Filters Used for Low-Level 

Environmental Surveillance Air Sampling, UCRL-JC-131844, University of California, 

Livermore, California.

DOE (U.S. Department of Energy), 1991, Environmental Regulatory Guide for Radiological Effluent 

Monitoring and Environmental Surveillance, DOE/EH-0173T, January. 

Gallegos, G. M., and K. A. Surano, 1992, Environmental Monitoring Plan, UCRL-ID-106132, 

Lawrence Livermore National Laboratory, Environmental Monitoring and Analysis Division, 

Livermore, California, April.

Gede, V. P., and P. D. Gildea, 1980, Tritium Practices Past and Present, UCRL-84318, Livermore, 

California, April 29.

LLNL (Lawrence Livermore National Laboratory), 1977, Environmental Monitoring at the Lawrence 

Livermore National Laboratory, 1976 Annual Report, UCRL-50027-76, University of California, 

Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 1980, Environmental Monitoring at the Lawrence 

Livermore National Laboratory, 1979 Annual Report, UCRL-50027-79, University of California, 

Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 1986, Environmental Monitoring at the Lawrence 

Livermore National Laboratory, 1985 Annual Report, UCRL-50027-85, University of California, 

Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 1987, Environmental Monitoring at the Lawrence 

Livermore National Laboratory, 1986 Annual Report, UCRL-50027-86, University of California, 

Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 1988, Environmental Monitoring at the Lawrence 

Livermore National Laboratory, 1986 Annual Report, UCRL-50027-87, University of California, 

Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 1989, Environmental Report for 1988, Lawrence 

Livermore National Laboratory, UCRL-50027-88, University of California, Livermore, 

California.

LLNL (Lawrence Livermore National Laboratory), 1992a, Environmental Monitoring Plan, UCRL-ID106132, University of California, Livermore, California.

LLNL (Lawrence Livermore National Laboratory), 1992b, Environmental Report for 1991, Lawrence 

Livermore National Laboratory, UCRL-50027-91, University of California, Livermore, 

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GLOSSARY

alpha radiation

Radiation consisting of charged particles identical with the isotope 4

He.

background radiation

Radiation received that is not associated with a worker’s occupation. This includes cosmic 

and terrestrial sources.

becquerel (Bq)

The derived SI unit of radioactivity equal to one disintegration per second.

beta radiation

Radiation consisting of electrons emitted spontaneously from the nuclei of certain radioactive 

elements.

deep dose equivalent

Dose equivalent at a depth of 1.0 cm in soft tissue.

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external and/or internal sources of radiation.

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, and 

neutrons from unstable nuclei.

rem

A unit of dose equivalent.

thermoluminescent dosimeter

A holder containing solid chips of material that when heated release the stored energy as light. 

The measurement of this light provides a measurement of absorbed dose.