ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 43

DOE Review Release 03/29/2010

Document Title:

Lawrence Livermore National Laboratory –

Occupational Environmental Dose

Document Number: ORAUT-TKBS-0035-4

Revision: 01

Effective Date: 03/16/2010

Type of Document: TBD

Supersedes: Revision 00

Subject Expert(s): Mutty M. Sharfi, Matthew H. Smith, and Stephen Spanos

Site Expert(s): N/A

Approval: Signature on File Approval Date: 03/10/2010

Ralph W. Kenning, Document Owner

Concurrence: Signature on File Concurrence Date: 03/10/2010

John M. Byrne, Objective 1 Manager

Concurrence: Keith A. McCartney Signature on File for Concurrence Date: 03/11/2010

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 03/11/2010

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 03/16/2010

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. 01 Effective Date: 03/16/2010 Page 2 of 43

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.

03/16/2010 01 Text revised to address comments from Objective 1 management 

and Objective 3. Adds Attributions and Annotations section. 

Incorporates formal internal and NIOSH review comments. 

Incorporates additional data on environmental neutron 

measurements and SEC Class recommended by NIOSH. 

Constitutes a total rewrite of the document. Training required: As 

determined by the Objective Manager. Initiated by Matthew H. 

Smith, Stephen Spanos and Ralph W. Kenning.
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 3 of 43

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

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

4.1.1 Purpose........................................................................................................................8

4.1.2 Scope...........................................................................................................................8

4.1.3 Special Exposure Cohort Petition Information for LLNL ................................................ 9

4.2 Internal Dose from Onsite Atmospheric Radionuclide Concentrations......................................9

4.2.1 Ambient Air Sample Collection Network ....................................................................... 9

4.2.1.1 Before 1959 ................................................................................................9

4.2.1.2 1959 through 1971 .................................................................................... 10

4.2.1.3 After 1971 ................................................................................................. 12

4.2.1.4 Purpose and Placement of Sampling Locations ........................................ 13

4.2.1.5 Radionuclides of Significance.................................................................... 13

4.2.2 Sampling and Analysis Methods................................................................................. 14

4.2.3 Estimation of Potential Annual Intakes from Airborne Radionuclides .......................... 15

4.2.3.1 Gross Alpha and Gross Beta before 1959................................................. 16

4.2.3.2 Gross Alpha and Gross Beta from 1959 to 1970 ....................................... 16

4.2.3.3 Tritium from 1952 to 1972 ......................................................................... 18

4.2.3.4 Tritium, Uranium, and Plutonium after 1970 .............................................. 19

4.3 Intakes from Tritium in Drinking Water from 1952 to 1978......................................................23

4.4 Ambient Radiation..................................................................................................................24

4.4.1 Gamma Radiation....................................................................................................... 24

4.4.2 Neutron Radiation....................................................................................................... 27

4.5 Attributions and Annotations ..................................................................................................32

References ........................................................................................................................................34

Glossary ............................................................................................................................................43
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LIST OF TABLES

TABLE TITLE

4-1 Main site ambient air sampling station locations.....................................................................

PAGE

10

4-2 Site 300 ambient air sampling station locations...................................................................... 12

4-3 Main site estimated air concentrations and annual median intakes via inhalation, 1952 to 

1958.......................................................................................................................................17

4-4 Site 300 estimated air concentrations and annual median intakes via inhalation, 1955 to 

1960.......................................................................................................................................17

4-5 Main site maximum annual median air concentrations and estimated intakes via 

inhalation, 1959 to 1970......................................................................................................... 17

4-6 Site 300 maximum annual median air concentrations and intakes via inhalation, 1961 to 

1970.......................................................................................................................................18

4-7 Main site maximum annual median air concentrations and intakes via inhalation, 1971 to 

2005.......................................................................................................................................21

4-8 Site 300 maximum annual median air concentrations and intakes via inhalation, 1971 to 

2005.......................................................................................................................................23

4-9 Maximum sitewide annual intakes (Bq/yr) from tritium via drinking water for main site and 

Site 300..................................................................................................................................24

4-10 Main site external gamma radiation........................................................................................ 28

4-11 Site 300 external gamma radiation......................................................................................... 30

4-12 Main site external neutron radiation ....................................................................................... 31

LIST OF FIGURES

FIGURE TITLE

4-1 Main site air particulate and tritium sampling locations...........................................................

PAGE

11

4-2 Site 300 and downtown Tracy air particulate and tritium sampling locations .......................... 12

4-3 Calculated annual median concentrations of 239+240Pu, 1986 to 2005 ..................................... 20

4-4 Main site gamma dosimeter locations, 2001........................................................................... 25

4-5 Site 300 and vicinity gamma dosimeter locations, 2001 ......................................................... 26

4-6 Main site neutron dosimeter locations .................................................................................... 28
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ACRONYMS AND ABBREVIATIONS

Bq becquerel

CEDE committed effective dose equivalent

CFR Code of Federal Regulations

Ci curie

d day

DCG derived concentration guideline

DOE U.S. Department of Energy

DOL U.S. Department of Labor

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EPA U.S. Environmental Protection Agency

ft foot

Ge(Li) lithium-drifted germanium

GSD geometric standard deviation

hr hour

in. inch

kCi kilocurie

L liter

LBNL Lawrence Berkeley National Laboratory

LLNL Lawrence Livermore National Laboratory 

m meter

MeV megaelectron-volt, 1 million electron-volts

MEI maximally exposed individual

MFP mixed fission products

mi mile

min minute

mrem millirem

mSv millisievert

nBq nanobecquerel

NESHAPs National Emissions Standards for Hazardous Air Pollutants

NIOSH National Institute for Occupational Safety and Health

ORAU Oak Ridge Associated Universities

PBq petabecquerel (1 × 1015 Bq)

POC probability of causation

SRDB Ref ID Site Research Database Reference Identification (number)

TBD technical basis document

TBq terabecquerel (1 × 1012 Bq)

TLD thermoluminescent dosimeter
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 6 of 43

U.S.C. United States Code

wk week

yr year

§ section or sections
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 7 of 43

4.1 INTRODUCTION

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2007):

] guidelines established under 

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

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

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

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

conventional structures

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

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
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4.1.1

Lawrence Livermore National Laboratory (LLNL) in Livermore, California, began operations in 1952 as 

the University of California Radiation Laboratory, which was a branch of what is now the Lawrence 

Berkeley National Laboratory (LBNL). It was later known as the Lawrence Radiation Laboratory and 

the Lawrence Livermore Laboratory before assuming its current name. The University of California 

operated LLNL for DOE and its predecessor agencies from 1952 through September 2007, when a 

partnership of the University of California, Bechtel Corporation, Babcock and Wilcox, the URS 

Corporation, and Battelle Memorial Institute assumed this role. Lawrence Livermore National Security 

has operated the site since then. Throughout its history, LLNL has processed and handled a variety 

of radionuclides, including uranium and transuranic elements, mixed fission products, and acceleratorproduced isotopes. This TBD discusses two sites: the main site, designated as Site 200, where 

research and development activities have taken place since 1952, and Site 300, where explosive 

testing experiments have occurred. Site 300 explosive testing included depleted and natural uranium, 

natural thorium, and tritium triggers. Site 300 began operations in 1955. This TBD addresses 

potential for internal dose from the breathing of airborne concentrations of radionuclides that have 

been released on the LLNL site. 

Purpose

The receptors of concern in this TBD are unmonitored workers, namely LLNL employees who did not 

wear external dosimetry or who were not monitored for internal exposures. Initially, LBNL, designated 

as Site 100, provided LLNL with beta/photon film dosimeters and processing, and PICs (Thompson 

1953). The practice of providing film and pocket ionization dosimeters to all workers has been in 

effect since March 1953 (Thompson 1953) and in 1958 film badges became part of the security badge 

(Nolan 1958), which effectively mandated that all workers wear their film dosimeters at all times. 

There is ongoing collaboration between these two laboratories. To provide the basis for estimating 

the environmental dose for years when monitoring did not occur, this TBD provides annual intakes 

and ambient external doses from 1952 (1955 for Site 300) to 2005 (the last year with publicly 

available data). 

4.1.2

This TBD documents historical practices at LLNL and provides information for the evaluation of 

environmental radiation data. It can serve as a supplement to, or substitute for, individual monitoring 

data.

Scope

Occupational environmental dose refers to dose from exposures workers received while on the site 

but outside the facilities at LLNL from elevated ambient radiation, facility discharges to the 

environment, and resuspension of radionuclides in soils. Effluents can result in 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 the estimated 

radiation doses as a result of ambient exposures at LLNL. Environmental measurements do not 

distinguish sources of emissions and, therefore, reflect air concentrations from nearby as well as 

distant sources. The estimates of emissions were useful in filling some gaps in measurement data 

and were critical to estimating exposures before the start of comprehensive and routine measurement 

data reports.

Section 4.2 of this TBD contains detailed information about the collection and analysis of air samples 

and the use of this information for dose reconstruction. There was a potential for ingestion of 

radioactive materials from resuspension of radioactive materials (Section 4.2) and for ingestion of 

water that might have contained tritium (Section 4.3). Section 4.4 describes the potential external 

doses from sources of radiation outside the process buildings.
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Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

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

4.1.3

Classes Added to the SEC

Special Exposure Cohort Petition Information for LLNL

NIOSH has determined, with concurrence from the Secretary of Health and Human Services (HHS), 

that MFP doses at LLNL cannot be reconstructed with sufficient accuracy between 1950 and 1973,

inclusive. For this reason, the following class of LLNL employees has been added to the Special 

Exposure Cohort (SEC) (Leavitt 2008). 

Employees of the DOE, its predecessor agencies, and DOE contractors or 

subcontractors who were monitored for radiation exposure while working at the 

Lawrence Livermore National Laboratory from January 1, 1950, through December 31, 

1973, for a number of workdays aggregating at least 250 workdays or in combination 

with workdays within the parameters established for one or more other classes of 

employees in the Special Exposure Cohort.

Class Recommended by NIOSH for addition to the SEC

NIOSH has subsequently determined that MFP doses for workers not monitored for radiation 

exposure at LLNL also cannot be reconstructed with sufficient accuracy between 1950 and 1973, 

inclusive. For this reason, NIOSH has recommended that the following class of LLNL employees be 

added to the SEC (NIOSH 2010). 

All employees of the Department of Energy, its predecessor agencies, and their 

contractors and subcontractors who worked at the Lawrence Livermore National 

Laboratory in Livermore, California from January 1, 1950 through December 31, 1973, 

for a number of work days aggregating at least 250 work days, occurring either solely 

under this employment or in combination with work days within the parameters 

established for one or more other classes of employees in the Special Exposure 

Cohort.

The NIOSH-recommended SEC class includes all workers during the SEC period, regardless of 

whether an individual worker was monitored for radiation exposure. Because of the identified dose 

reconstruction infeasibility, all dose reconstructions for all workers having employment during the SEC 

period are considered partial dose reconstructions. If monitoring data are available for workers 

included in the SEC, dose is to be assigned as appropriate based on such data; however, such dose 

reconstructions are still considered partial dose reconstructions because the HHS has determined that 

exposure to mixed fission and activation products during the SEC period cannot be bounded.

4.2 INTERNAL DOSE FROM ONSITE ATMOSPHERIC RADIONUCLIDE 

CONCENTRATIONS

4.2.1

4.2.1.1 Before 1959

Ambient Air Sample Collection Network

A review of records indicates that the earliest available environmental moniotoring data for the LLNL

main site was 1959 (Beaufait 1961), and for Site 300 was 1961 (LRL 1962). Before the 1980s LLNL’s 

almost exclusive mission was weapons research. Therefore, it was assumed that the radionuclide 

emission rate would be proportional to the weapons-related activity at the site from design and 
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fabrication to analysis of samples that were retrieved after a test. The peak for testing at LLNL was 

from 1962 through 1971 with an average of 31 tests per year (DOE 2000). From 1953 through 1961 

LLNL sponsored an average of 7 tests per year (DOE 2000). During the entire period of LLNLsponsored weapons testing from 1953 through 1992 the average was 14.5 per year. 

It was assumed that all activity that was detected at the monitoring stations was from LLNL activities. 

This assumption is conservative because this was during a time of high fallout due to atmospheric 

testing, and in many cases offsite locations had higher activity than onsite locations due to the fallout. 

Therefore, for the early years before monitoring, dose reconstruction should use the highest 

measured results from the later years in which monitoring occurred. Section 4.2.3.1 provides 

estimated intakes based on the highest measured activity, despite location, as being favorable to 

claimants. For workers that should have been monitored, intakes based on coworker data might be 

more appropriate than environmental data.

4.2.1.2 1959 through 1971

Air samples were collected in 7 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 7-day period on HV-70 paper 4 in. wide by 9 in. long (LRL 1962). The other air 

sampling stations were at a variety of distances from along the perimeter to as far as 5 mi from the 

perimeter (LRL 1962). Table 4-1 lists the seven onsite sampling stations this analysis used for 

estimates of occupational environmental exposures, and Figure 4-1 shows their locations as black 

dots. The SALV and MESQ locations were established as early as 1961 for perimeter air monitoring 

(LRL 1962). 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 (MEI) for LLNL. 

Table 4-1. Main site ambient air sampling station locations (Gallegos et al. 2002).

Designation Location

Direction from 

center of LLNL

COW 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 MEI

SE

At Site 300, six air sampling stations were established in 1962, each within the site perimeter. The air 

sampling equipment at Site 300 was equivalent to the equipment at the main site. Table 4-2 lists 

these stations, and Figure 4-2 shows their locations. The Bunker 801E, ECP, EOBS, GOLF, WCP, 

and WOBS locations were established in 1962. As stated above, this was during a time of high fallout

due to atmospheric testing, and in many cases offsite locations had higher activity than onsite 

locations due to the fallout. Therefore, Section 4.2.3.1 presents intakes estimated from the highest 

measured activity, despite location, as being favorable to claimants.
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Figure 4-1. Main site air particulate and tritium sampling locations (Gallegos et al. 2002).
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Table 4-2. Site 300 ambient air sampling station locations (Gallegos et al. 2002).

Designation Location

Direction from 

center of Site 300

COHO Near south perimeter of facility; representative of MEI 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

Figure 4-2. Site 300 and downtown Tracy air particulate and tritium sampling locations (Gallegos et 

al. 2002).

4.2.1.3 After 1971

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

perimeter and other locations inside the site (Gallegos et al. 1992). LLNL collected and analyzed air 

samples for gross alpha and beta radiations, such as 3

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

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

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

City of Tracy (Gallegos et al. 1992). Tables 4-1 and 4-2 list the stations, and Figures 4-1 and 4-2 

show their locations; black triangles represent tritium monitoring stations. As stated above, this was 
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 13 of 43

during a time of high fallout due to atmospheric testing and in many cases offsite locations had higher 

activity than onsite locations due to the fallout. The highest activity is presented, despite location, as 

being favorable to claimants.

4.2.1.4 Purpose and Placement of 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 reasonably be detected regardless 

of local meteorology (Griggs and Buddemeier 1986). This demonstrated compliance with DOE 

derived concentration guidelines (DCGs) 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 (Griggs and 

Buddemeier 1986). This TBD considers only those ambient air monitoring locations inside the LLNL 

security fence for worker intake unless a higher value was reported for an offsite monitoring location, 

in which case that data is reported to be favorable to claimants. Table 4-1 lists these locations.

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

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

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

as the sitewide 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 from historical operations that included solar evaporators for plutonium-containing 

liquid waste.

The Site 300 air particulate monitoring network included eight sampling units that were 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 sitewide MEI location 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 (Althouse et al. 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 main 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). 

4.2.1.5 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 from 1959 through 

1970 and for specific isotopes afterward that represented more than 90% of the LLNL radioactive 

materials inventory. It analyzed air samples specifically for the presence of tritium, 238

Pu, and 

239+240Pu, and isotopes of uranium. From 1959 through 1970, dose reconstructors should select the 

assigned dose from the higher of two calculated doses: one using the gross alpha concentration to 

calculate the presence of 239Pu, and a second using 234U. The gross beta concentration should be 

assumed to result from 90Sr [1]. 

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

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

uranium indicated the analysis of 238U and 235U. For estimates based on this TBD, the dose 

reconstructor should assign the dose that is associated with 234U for depleted uranium. Through 
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1971, dose reconstruction should use the gross alpha concentration for Site 300 to calculate the 

presence of 234U; it is assumed that the maximum radiation dose is assigned as 100% 234U versus a 

varying percentage of 234U and 235U [2]. Data are listed in the tables in Section 4.2.3.2. Assume the 

gross beta concentration is 234Th [3].

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, 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 

(Gallegos et al. 1992). 

Sampling and Analysis Methods

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 on air 

filters were used only as trend indicators; specific radionuclide analysis occurred for plutonium and 

uranium dependent on location. All analytical results were reported as a measured concentration per 

volume of air. When activity was less than the minimum detectable 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 EPA-certified standards. 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 in the counting equipment. The analytical laboratory 

reported the actual instrumentation values, which included negative results when background 

measurements were higher than those for the filters (DOE 1991).

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

offsite locations near the Livermore site, and at Site 300. From 1959 to 1964, air samples were 

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

sample flow rate of about 4 ft3

/min of air (LRL 1964). As discussed in Section 4.2.1.1, no monitoring 

data was found for the years before 1959 for the main site, and 1961 for Site 300. The samples were 

collected for 7 days. After a 4-day delay for decay of radon and 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. 

After gamma counting, the perimeter filters were grouped by sampling location. 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 perimeter were sampled 

(Hughey 1965). The six samplers on the Laboratory perimeter used 0.052 m2 Whatman-41 cellulose 

filters. The particulates were collected on these filters using an average air flow rate of 700 L/min 

(Lindeken et al. 1978). Tritium was collected with columns packed with silica gel using 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 assurance 

samplers operated in parallel with the permanent sampler at each site; samples from these duplicates 

were analyzed to confirm results (Lindeken et al. 1978). 
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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 including 235U, 238U, 40K, 226Ra, 228Ra, and 228Th (Althouse 1998). In fact, the amount of these 

naturally occurring isotopes in the filters is often greater than the amount the filter captures from the 

air. The filters are 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 (Althouse et al. 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 analyses were 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 MEI (from the 

total particulate) was below the 10-mrem committed effective dose equivalent (CEDE) (Althouse et al. 

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 of the 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 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 samples were processed to confirm the precision of the 

analytical results obtained from the samplers. A duplicate 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 

collection trip. The trip blanks and duplicates were processed in the same manner as that for the 

routine samples and analyzed for the same radiological parameters. 

4.2.3

This section discusses estimation of annual intakes from 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 1959 to 

2005.

Estimation of Potential Annual Intakes from Airborne Radionuclides

2

 2 The References section lists all directly cited references, and then separately lists all environmental reports in date order.

 The analysis for this TBD collected measurements from the air monitoring locations in 

Tables 4-1 and 4-2 from these documents and analyzed them in accordance with the collection and 

analysis methods Section 4.2.2 describes. If data were not available or additional information was 

necessary, these documents were supplemented by information from DOE or the University of 

California. 
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 16 of 43

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

air concentrations over the history of LLNL. Over the past 40 years of monitoring, concentrations of 

airborne radioactive materials have decreased by more than 3 orders of magnitude. These factors 

include: 

• Improvements in monitoring methods that reduced minimum detectable concentrations;

• Reductions in releases of naturally occurring radioactive material from a nearby fossil-fuel 

generating plant, which has improved its environmental controls over the years; and 

• Reductions in the effects of atmospheric weapons testing including radioactive decay of 

fallout.

The method for calculation of the annual intakes from onsite atmospheric radionuclide concentrations 

was applied directly to available air sampling data. Air sampling measurement error and uncertainty 

were evaluated using the maximum reported value 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 the estimates of airborne concentrations at specific locations around the 

LLNL site using traditional transport modeling approaches:

• Number of release points;

• Characteristics of the release points, which include stacks, vents, and other emission sources;

• Limited number of air sampling locations;

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

• Density and configurations of buildings at the site.

Section 4.2.3.1 and 4.2.3.2 provide information on gross alpha and gross beta inhalation exposures 

for the periods from 1952 to 1958 for the main site, and from 1955 to 1960 for Site 300 and from 1959 

to 1970 for the main site, and from 1961 to 1970 for Site 300, respectively; the estimates for the 

earlier period are based on the data from the later period. Section 4.2.3.3 discusses tritium exposures 

for the period from 1952 to 1972, before tritium monitoring began. Section 4.2.3.4 provides estimated 

inhalation exposures based on air monitoring for tritium, plutonium, and uranium for 1971 to 2005.

4.2.3.1 Gross Alpha and Gross Beta before 1959

For the main site before 1959, and for Site 300 before 1961, environmental air monitoring samples 

were not available. The estimated air concentrations and potential intakes of gross alpha and gross 

beta activity before 1959 for the main site, and 1961 for Site 300 are listed in Tables 4-3 and 4.4. The 

analysis assumed a breathing rate of 2,400 m3

/yr based on 20 L/min for 2,000 hr/yr, and a geometric 

standard deviation (GSD) for these data of 3 [4]. This data is based on using the highest of the 

measured data for years during which environmental monitoring occurred. All environmental internal 

doses assigned prior to 1959 are considered partial dose assessments due to the LLNL SEC for MFP.

4.2.3.2 Gross Alpha and Gross Beta from 1959 to 1970

Data from the environmental reports were available for 1959 to 1970 for the main site and for 1961 to 

1970 for Site 300. The air monitoring data provided no isotopic analysis; the data were reported only 

for the gross alpha and gross beta radiations. Table 4-5 list air concentrations and estimated intakes 

for the LLNL site, and Table 4-6 lists the values for Site 300. 
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 17 of 43

Table 4-3. Main site estimated air concentrations and annual 

median intakes via inhalation, 1952 to 1958.

Year

Gross alpha 

(Pu-239 or U-234)

Gross beta 

(Sr-90)

Air 

concentration

(Bq/m3)

Annual 

intake

(Bq/yr)

Air 

concentration

(Bq/m3)

Annual 

intake

(Bq/yr)

1952 1.6E-3 3.8 6.7E-1 1,610

1953 1.6E-3 3.8 6.7E-1 1,610

1954 1.6E-3 3.8 6.7E-1 1,610

1955 1.6E-3 3.8 6.7E-1 1,610

1956 1.6E-3 3.8 6.7E-1 1,610

1957 1.6E-3 3.8 6.7E-1 1,610

1958 1.6E-3 3.8 6.7E-1 1,610

Table 4-4. Site 300 estimated air concentrations and annual 

median intakes via inhalation, 1955 to 1960.

Year

Gross alpha 

(U-234)

Gross beta 

(Th-234)

Air 

concentration 

(Bq/m3)

Annual 

intake 

(Bq/yr)

Air 

concentration 

(Bq/m3)

Annual 

intake 

(Bq/yr)

1955 6.7E-3 16 2.4E+0 5760

1956 6.7E-3 16 2.4E+0 5760

1957 6.7E-3 16 2.4E+0 5760

1958 6.7E-3 16 2.4E+0 5760

1959 6.7E-3 16 2.4E+0 5760

1960 6.7E-3 16 2.4E+0 5760

Table 4-5. Main site maximum annual median air 

concentrations and estimated intakes via inhalation, 1959 to 

1970.

Year

Gross alpha 

(Pu-239 or U-234)

Gross beta 

(Sr-90)

Air 

concentration

(Bq/m3)

Annual

intake

(Bq/yr)

Air 

concentration

(Bq/m3)

Annual 

intake

(Bq/yr)

1959 7.4E-4 1.8 2.7E-1 648

1960 8.5E-4 2.0 6.7E-1 1,610

1961 1.6E-3 3.8 4.1E-1 984

1962 1.7E-4 0.4 9.0E-3 22

1963 1.2E-4 0.3 1.3E-1 312

1964 1.2E-4 0.3 2.4E-2 58

1965 3.7E-5 0.1 6.5E-3 16

1966 3.7E-5 0.1 2.6E-3 6

1967 2.9E-5 0.07 4.7E-3 11

1968 3.7E-5 0.1 4.4E-3 10

1969 3.7E-5 0.1 3.8E-3 9

1970 3.7E-5 0.1 5.6E-3 14
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 18 of 43

Table 4-6. Site 300 maximum annual median air 

concentrations and intakes via inhalation, 1961 to 1970.

Year

Gross alpha 

(U-234)

Gross beta 

(Th-234)

Air 

concentration 

(Bq/m3)

Annual 

intake

(Bq/yr)

Air 

concentration 

(Bq/m3)

Annual 

intake 

(Bq/yr)

1961 6.7E-3 16 2.6E-3 6.2

1962 1.9E-2 44 1.4E-1 336

1963 3.4E-3 8 2.4E+0 5760

1964 1.3E-3 3 1.9E-1 456

1965 1.3E-4 0.3 2.4E-2 58

1966 2.0E-4 0.5 2.9E-2 70

1967 1.9E-4 0.5 1.6E-1 384

1968 3.7E-5 0.09 2.5E-2 60

1969 7.4E-5 0.2 3.9E-2 94

1970 4.1E-4 1 3.4E-2 82

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

The presence of beta radiation indicates the progeny of uranium. 

All environmental internal doses assigned from 1952 to 1970 are considered partial dose 

assessments due to the LLNL SEC for MFP. 

4.2.3.3 Tritium from 1952 to 1972

Environmental air monitoring samples were not analyzed for tritium before 1972. Most of the released 

tritium was airborne with the dominant source being the Tritium Facility (Building 331, previously 

Building 172), which began operation in 1952. The total activity of tritium released was reported to be 

approximately 28 PBq (760 kCi) over the more than 40 years of LLNL history (Tate et al. 1999). Most 

of this activity was released in two events: 13 PBq (350 kCi) on January 21, 1965, beginning at 3:30 

p.m., and 11 PBq (300 kCi) on August 6, 1970, beginning at 6:14 a.m. Both releases were in the form 

of hydrogen gas and were released from a 30-m-tall stack over a period of less than 1 hr (DOE 1982). 

Dispersion modeling estimated air concentrations of tritium to range from 3.2 × 102 to 5.2 × 108 Bq/m3

over the period of the release within 0.5 mi of the stack. Potential intakes of tritium were estimated for 

the release to range from 300 to 6 × 108 Bq with a corresponding estimated radiation dose of less 

than 8 mrem; the 95th-percentile dose was calculated to be less than 19 mrem. 

Assuming the remaining 4 PBq was released over the 40-year history as routine discharges 

(exclusive of the releases in 1965 and 1970), an estimate of potential exposure can be derived for the 

period for which no environmental data are available. To be favorable to the claimant, it is assumed 

that the 4 PBq of tritium was released before 1972 and distributed equally for each year. That is, it is 

assumed that 200 TBq were discharged each year [5]. A study that reviewed the release of tritium 

reported a range of 40 to 280 TBq/yr (1 to 7.5 kCi) before 1992 and between 3 and 12 TBq (80 and 

320 Ci) since 1992 (Tate et al. 1999). 

Using the dispersion calculations from 1970, the estimated annual release of 200 TBq/yr, and the 

same weather conditions that existed during the 1970 release, the estimated concentration of tritium 

ranges from 5.8 to 2 × 106 Bq/m3 for 1952 to 1971. It is unreasonable to assume that a worker stayed 

in the exact locations of maximum exposure (immediately downwind of the Building 331 stack and at 

the points of maximum impact) for the entire exposure period of 20 years. Therefore, it is not 

reasonable to assume that the maximum concentration represents the likely potential exposure. The 
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 19 of 43

estimated concentration of 1.7 × 104 Bq/m3 was derived for the maximum site boundary location and 

is more likely to represent the average potential exposure to a single worker at LLNL for the exposure 

period. The value of 1.7 × 104 Bq/m3 was calculated using the concentration at the site boundary 

0.5 mi northeast (about 40°) from the Building 331 release point for the 1970 release (ATSDR 2001). 

The concentration at this point was divided by the total 1970 11 PBq (300 kCi) release; the resultant 

value was then multiplied by the estimated annual release of 200 TBq/yr. The 40° direction was 

chosen because it closely reflects (within 5°) to the predominant southwest wind direction for LLNL 

(Althouse et al. 2001). 

For purposes of an estimated radiation dose that is favorable to claimants, an intake of 3 x 107 Bq 

(1.7 × 104 Bq/m3

) of tritium as HTO should be used for each year from 1952 to 1971. This value is 

considered favorable to claimants because only approximately one-half of the total activity of tritium 

releases occurred during this period (excluding the 1965 and 1970 events). In addition, the airborne 

concentration of 3.14 in becquerel per cubic meter per released curie in a year was compared with the 

onsite measurements (rather than the perimeter measurements) that were made in the immediate 

vicinity of the diffuse sources of tritium beginning in 1991, such as Building 331 (Gallegos et al. 1992). 

The value of 1.7 × 104 Bq/m3 represents 3.14 Bq/m3 per released curie during the year. The highest 

single measured value adjacent to an onsite source for 1991 through 2000 was 0.46 Bq/m3 per 

released curie, which occurred in 1995; this value is approximately 20 times higher than the maximum 

median onsite measurement for that year (Harrach et al. 1996a,b). The highest maximum tritium 

median value was 0.15 Bq/m3 per released curie, which occurred in 1998. Further favorability to 

claimants arises because all the tritium is considered to be HTO even though a large fraction of the 

emissions were in the form of hydrogen (tritium) gas, and HTO has a higher dose conversion factor 

than tritium gas [6]. 

4.2.3.4 Tritium, Uranium, and Plutonium after 1970

Estimates of offsite doses to members of the public from the 1971 to 2005 environmental reports (see 

the References section) indicated that the potential internal dose from airborne releases to LLNL 

workers should be relatively low (10-mrem CEDE or less). 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 4-7 lists main site concentrations for 3

H, 239+240Pu, and 234U (3H not reported for 1971). This 

TBD assumed that a worker inhaled 20 L of air per minute for 2,000 hr/yr or approximately 2.4 × 106 L 

of air per year (2,400 m3

/yr). The concentration of tritium ranged from 0.04 to 3 Bq/m3; the potential 

for exposure to tritium ranged from 144 to 11,200 Bq/yr. For tritium, the total uptake by the body was 

increased by 50% to account for absorption by the skin in addition to the inhalation pathway (Gede 

and Gildea 1980). Figure 4-3 shows how the concentration of 239+240Pu decreased from 1986 through 

2005. The concentration dropped by more than 20 times over the 20 years of sampling. 
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 20 of 43

Figure 4-3. Calculated annual median concentrations of 239+240Pu, 1986 to 2005 (Gallegos et al. 

2002).
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 21 of 43

Table 4-7. Main site maximum annual median air concentrations and intakes via 

inhalation, 1971 to 2005.

a

Year

Tritium Pu-239+240 U-234b

Air 

concentration 

(Bq/m3)

Annual 

intake 

(Bq/yr)c

Air 

concentration 

(Bq/m3)

Annual 

intake 

(Bq/yr)

Air 

concentration 

(Bq/m3)

Annual 

intake 

(Bq/yr)

1971 (d. e) (d, e) 4.1E-6 9.8E-3 2.2E-6 5.3E-3

1972 1.4 5.0E+3 2.2E-6 5.3E-3 3.3E-6 7.8E-3

1973 1.4 5.0E+3 2.7E-4 6.5E-1 1.3E-6 3.1E-3

1974 1.7 6.2E+3 3.1E-6 7.4E-3 3.2E-6 7.6E-3

1975 1.7 6.2E+3 1.3E-6 3.1E-3 1.3E-6 3.1E-3

1976 3.0 1.1E+4 1.0E-6 2.4E-3 1.1E-6 2.6E-3

1977 2.2 8.0E+3 1.6E-6 3.8E-3 1.6E-6 3.8E-3

1978 2.0 7.2E+3 2.0E-6 4.8E-3 1.1E-6 2.6E-3

1979 1.6 5.7E+3 8.5E-7 2.0E-3 1.1E-6 2.6E-3

1980 1.4 5.0E+3 5.2E-7 1.2E-3 9.2E-7 2.2E-3

1981 1.6 5.7E+3 5.0E-7 1.2E-3 1.1E-6 2.6E-3

1982 1.4 5.0E+3 5.0E-7 1.2E-3 1.1E-6 2.6E-3

1983 1.1 3.9E+3 8.9E-7 2.1E-3 1.9E-6 4.7E-3

1984 1.3 4.7E+3 5.0E-7 1.2E-3 1.1E-6 2.6E-3

1985 1.0 3.6E+3 2.2E-7 5.3E-4 1.1E-6 2.6E-3

1986 0.9 3.3E+3 1.5E-7 3.6E-4 1.3E-6 2.8E-3

1987 1.7 6.2E+3 1.9E-7 4.6E-4 1.3E-6 3.1E-3

1988 1.3 4.7E+3 3.3E-8 7.9E-5 2.7E-6 6.5E-3

1989 0.9 3.3E+3 1.4E-7 3.4E-4 1.1E-6 2.8E-3

1990 0.5 1.8E+3 1.5E-7 3.6E-4 6.4E-6 1.5E-2

1991 0.7 2.6E+3 7.0E-8 1.7E-4 1.9E-6 4.6E-3

1992 0.2 7.2E+2 6.8E-8 1.6E-4 1.1E-6 2.6E-3

1993 0.2 7.2E+2 2.7E-8 6.5E-5 8.7E-7 2.1E-3

1994 0.1 3.6E+2 3.4E-8 8.2E-5 6.8E-7 1.6E-3

1995 0.07 2.6E+2 3.5E-8 8.4E-5 6.9E-7 1.7E-3

1996 0.2 7.2E+2 2.4E-8 5.8E-5 7.7E-7 1.9E-3

1997 0.1 3.6E+2 8.8E-9 2.1E-5 6.6E-7 1.6E-3

1998 0.09 3.3E+2 8.6E-9 2.1E-5 4.6E-7 2.1E-3

1999 0.09 3.3E+2 5.5E-9 1.3E-5 8.9E-7 2.1E-3

2000 0.05 1.8E+2 9.1E-9 2.2E-5 1.3E-8 3.1E-5

2001 0.04 1.4E+2 9.1E-9 2.2E-5 1.3E-8 3.9E-5

2002 0.64 1.5E+3 6.1E-9 1.5E-5 1.5E-6 5.3E-8

2003 2.8 6.7E+3 4.8E-9 1.2E-5 4.5E-7 1.1E-3

2004 0.06 1.4E+2 4.6E-9 1.1E-5 2.9E-7 7.0E-4

2005 0.5 1.2E+3 1.5E-9 3.6E-6 2.0E-7 4.8E-4

a. The data are the maximum of the median measured value for a specific year from the available stationary 

monitoring locations.

b. The actual concentration for this isotope was not reported. Isotopes of uranium were summed and 

reported as U-234.

c. Assume total intake is the sum of inhalation and skin absorption. Total intake is 1.5 times the inhalation 

intake.

d. Tritium was not reported for 1971.

e. Per Section 4.2.3 an intake of 3 x 107 Bq/yr (1.7 × 104 Bq/m3

) of tritium as HTO should be assumed for

1952 to 1971.

As Section 4.2.1.5 describes, the isotopes of significance at LLNL were 239+240Pu and 234U, which 

decay by alpha radiation. Isotopes that decay by beta radiation were determined to be progeny of 

these two parent isotopes. 
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 22 of 43

Tables 4-7 and 4-8 indicate the potential for exposure to inhalation of airborne radionuclides at the 

main site and Site 300 respectively. The Site 300 tritium doses are the same as those for the main 

site, which is favorable to claimants because the Site 300 values are consistently less than 10% of the 

main site values. Post 1970, only alpha emitting radionuclides are listed. This is because the gross 

beta annual intakes post 1970 trend down to where the resulting internal doses are less than 0.001 

rem. Radionuclides contributing doses less than 0.001 rem, are not considered significant. 

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 

through 4-8 list these data and demonstrate that the chosen values are reasonable and favorable to 

claimants. Using the maximum airborne radionuclide concentration in a year (either from actual data 

or an estimated value), annual intakes that are favorable to claimants for the radionuclides of concern 

were derived using an assumed individual annual respiration rate of 2,400 m3

/yr.

All environmental internal doses assigned from 1971 through 1973 are considered partial dose 

assessments due to the LLNL SEC for MFP. 
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 23 of 43

Table 4-8. Site 300 maximum annual median air concentrations and intakes via 

inhalation, 1971 to 2005.

Year

Tritiuma Pu-239+240b U-234c

Air 

concentration 

(Bq/m3)

Annual 

intake 

(Bq/yr)

Air 

concentration 

(Bq/m3)

Annual 

intake

(Bq/yr)

Air 

concentrati

on 

(Bq/m3)

Annual 

intake

(Bq/yr)

1971 (d, e) (d, e) 4.1E-6 9.8E-3 1.1E-5 2.6E-2

1972 1.4 5.0E+3 2.3E-6 5.5E-2 2.9E-5 7.0E-2

1973 1.4 5.0E+3 2.1E-6 5.0E-3 9.1E-6 2.2E-2

1974 1.7 6.2E+3 4.1E-6 9.8E-3 9.3E-6 2.2E-3

1975 1.7 6.2E+3 2.1E-6 5.0E-3 6.2E-6 1.5E-2

1976 3.0 1.1E+4 6.2E-6 1.5E-2 2.0E-6 4.8E-3

1977 2.2 8.0E+3 8.6E-6 2.1E-2 1.5E-6 3.6E-3

1978 2.0 7.2E+3 1.2E-5 3.0E-2 2.8E-6 6.7E-3

1979 1.6 5.7E+3 6.2E-6 1.5E-2 1.6E-6 3.7E-3

1980 1.4 5.1E+3 8.5E-7 2.1E-3 1.8E-6 4.3E-3

1981 1.6 5.7E+3 8.6E-5 2.1E-1 1.7E-6 4.0E-3

1982 1.4 5.1E+3 8.5E-7 2.1E-3 1.8E-6 4.3E-3

1983 1.1 3.9E+3 2.0E-5 4.9E-2 7.4E-8 1.8E-4

1984 1.3 4.7E+3 8.5E-7 2.1E-3 1.8E-6 4.3E-3

1985 1.0 3.6E+3 3.0E-7 7.2E-4 1.1E-6 2.1E-3

1986 0.9 3.3E+3 7.0E-8 1.7E-4 8.3E-7 2.0E-3

1987 1.7 6.2E+3 5.9E-8 1.4E-4 2.8E-6 6.7E-3

1988 1.3 4.7E+3 3.9E-7 9.6E-4 2.7E-6 6.5E-3

1989 0.9 3.3E+3 3.5E-8 8.4E-5 2.5E-6 5.9E-3

1990 0.5 1.8E+3 1.2E-8 2.8E-5 2.5E-6 5.8E-3

1991 0.7 2.6E+3 7.8E-8 1.8E-4 3.4E-6 8.0E-3

1992 0.2 7.2E+2 1.4E-8 3.4E-5 3.0E-6 7.2E-3

1993 0.2 7.2E+2 7.3E-9 1.8E-5 1.9E-6 4.6E-3

1994 0.1 3.6E+2 7.3E-9 1.8E-5 1.9E-6 4.6E-3

1995 0.07 2.6E+2 1.2E-8 2.9E-5 1.7E-7 4.1E-4

1996 0.2 7.2E+2 1.0E-8 2.4E-5 2.4E-6 5.8E-3

1997 0.1 3.6E+2 1.7E-8 4.1E-5 2.2E-6 5.3E-3

1998 0.09 3.3E+2 1.6E-8 3.8E-5 7.0E-6 1.7E-2

1999 0.09 3.3E+2 8.7E-9 2.1E-5 2.2E-6 5.3E-3

2000 0.05 1.8E+2 1.9E-8 4.6E-5 3.9E-6 1.4E-2

2001 0.04 1.4E+2 9.8E-9 2.3E-5 8.8E-6 2.1E-2

2002 0.64 1.5E+3 1.0E-8 2.4E-5 1.2E-6 2.9E-3

2003 2.8 6.7E+3 1.8E-9 4.3E-6 7.3E-7 1.8E-3

2004 0.06 1.4E+2 4.5E-9 1.1E-5 6.3E-7 1.5E-3

2005 0.5 1.2E+3 4.3E-9 1.0E-5 2.0E-6 4.8E-3

a. From Table 4-7.

b. Originating from fallout (Gallegos 1998).

c. Isotopes of uranium were summed and reported as U-234.

d. Tritium was not reported for 1971.

e. Per Section 4.2.3 an intake of 3 x 107 Bq/yr (1.7 × 104 Bq/m3

) of tritium as HTO should be assumed for 

1955 to 1971.

4.3 INTAKES FROM TRITIUM IN DRINKING WATER FROM 1952 TO 1978

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

LLNL (see the References section). Several water sources were sampled and analyzed; five were 

identified as sources of drinking water and are designated BELL, GAS, PALM, ORCH, and TAP. The 

water from these locations was sampled at least once per year. In addition, LLNL sampled a site 

called POOL, which was the onsite swimming pool. The median activity in the drinking water was 
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 24 of 43

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 (Gallegos et al. 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 550 to 220,000 Bq/yr. Table 4-9 lists the potential for ingestion of tritium in drinking water 

over time. The values in the table before 1966 were selected from the maximum concentration that 

was reported for the period from 1966 to 2005. No data were available for drinking water at Site 300; 

this TBD assumed that the concentration of tritium was the same as that at LLNL.

Table 4-9. Maximum sitewide annual intakes (Bq/yr) from tritium via drinking water 

for main site and Site 300c

. 

Year

Water 

concentration 

(Bq/L)

Annual intakea

(Bq/yr) Year

Water 

concentration 

(Bq/L)

Annual 

intakea

(Bq/yr)

1952 200b 2.2E+5 1979 11 1.2E+4

1953 200b 2.2E+5 1980 26b 2.9E+4

1954 200b 2.2E+5 1981 26b 2.9E+4

1955 200b 2.2E+5 1982 26b 2.9E+4

1956 200b 2.2E+5 1983 26b 2.9E+4

1957 200b 2.2E+5 1984 26b 2.9E+4

1958 200b 2.2E+5 1985 7.4 8.1E+3

1959 200b 2.2E+5 1986 7.4 8.1E+3

1960 200b 2.2E+5 1987 3.7 4.1E+3

1961 200b 2.2E+5 1988 6.6 7.2E+3

1962 200b 2.2E+5 1989 7b 7.7E+3

1963 200b 2.2E+5 1990 7b 7.7E+3

1964 200b 2.2E+5 1991 4.5 5.0E+3

1965 200b 2.2E+5 1992 4.4 4.8E+3

1966 185 2.0E+5 1993 1.0 1.1E+3

1967 185 2.0E+5 1994 2.8 3.1E+3

1968 185 2.0E+5 1995 0.8 8.8E+2

1969 185 2.0E+5 1996 2.1 2.3E+3

1970 200 2.2E+5 1997 2.4 2.6E+3

1971 200b 2.2E+5 1998 1.2 1.3E+3

1972 200b 2.2E+5 1999 2.2 2.4E+3

1973 200b 2.2E+5 2000 2.0 2.2E+3

1974 200b 2.2E+5 2001 3.0 3.3E+3

1975 200b 2.2E+5 2002 4.8 5.3E+3

1976 26 2.9E+4 2003 5.7 6.3E+3

1977 26b 2.9E+4 2004 3.1 3.4E+3

1978 26b 2.9E+4 2005 0.5 5.5E+2

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

b. The data for this year were not provided by LLNL. The maximum concentration was reported in 

1970. Therefore the 1970 maximum annual tritium intake should be used for the years where no 

data was reported.

c. Site 300 started in 1955. Tritium doses shold be assigned from 1995 and beyond.

4.4 AMBIENT RADIATION

4.4.1

From 1976 to the present, thermoluminescent dosimeters (TLDs) have been used on the main site to 

determine ambient external 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 

Gamma Radiation
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 25 of 43

the monitoring stations at the main site and Site 300, respectively. The data for this analysis were 

summarized from the annual environmental reports for 1989 to 2005 (see the References list). 

Figure 4-4. Main site gamma dosimeter locations, 2001 (Gallegos et al. 2002).
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 26 of 43

Figure 4-5. Site 300 and vicinity gamma dosimeter locations, 2001 (Gallegos 

et al. 2002).

Observations from earlier environmental reports, which were based on surveys with portable 

instruments, state, “0.02 mrem/h at all locations.” The instrumentation might be considered primitive 

by today’s standards, and the site did not employ continuous monitoring, so these results have been 

overshadowed by recent and more reliable observations. 

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. 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 that was 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 values for 

State of California 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.

The locations of the monitoring points from which data were summarized in this section add 

uncertainty to the results. The monitoring points, as stated above, have been located around the 
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 27 of 43

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 

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. The environmental reports gave the results as cumulative exposures in 

millirem per year based on 8,760 hours of exposure, which is the total number of hours in 365 days. 

External dose for some years was not available. For the main site, the maximum external dose was 

reported in 1975; the maximum site average was reported in 1972. Therefore, the 1972 and 1975

dose values were used respectively for the years noted. For Site 300, the maximum value was noted 

in 1994; this value was used for the average for years before 1989 when monitoring began. 

Tables 4-10 and 4-11 provide, for the main site and Site 300, respectively, average and maximum 

annual exposures for the 8,760-hr/yr exposure period. The yearly exposure values provided in Tables 

4-10, and 4-11 should be adjusted in accordance with ORAUT-PROC-0060, Occupational Onsite 

Ambient Dose Reconstruction for DOE Sites (ORAUT 2006).

4.4.2

Data that describe ambient neutron measurements on the main site were found in site environmental 

radiation reports beginning in 1972 (Gudiksen et al. 1973). From 1972 through 1987, elevated 

ambient neutron dose was measured at the perimeter fence near Building 212. The source of these 

elevated readings was the Insulating Core Transformer accelerator, which began operations in 1966 

in Building 212 and produced 14-MeV neutrons (Olson 1974). Figure 4-6 shows the locations of the 

environmental neutron dosimeters. Stations 5 and 3 were typically the maximum value locations. 

Neutron Radiation

Table 4-12 lists the maximum and available average ambient neutron values for the main site. Only 

maximum values were available from the reports for 1972 to 1980. Average and maximum values 

were available for 1981 to 1993. Again, the values from the original data were for an 8,760-hr/yr 

exposure period. The yearly exposure values provided in Table 4-12 should be adjusted in 

accordance with ORAUT-PROC-0060, Occupational Onsite Ambient Dose Reconstruction for DOE 

Sites (ORAUT 2006).

Potential ambient neutron dose is captured by application of the neutron-to-photon ratio to gamma 

dose (applied pre-1969) or by using measured and missed neutron dose data for the period from 

1969 to the present. Considering that all LLNL workers were using neutron sensitive TLDs beginning 

at least in 1969, any significant neutron dose should have been detected and measured by these 

dosimeters (Unruh ca. 1969). Therefore, it is not necessary to consider adding additional ambient 

neutron dose for monitored workers.

If dosimetry data are not available for a worker, and if the worker was likely located in the area of the 

LLNL campus near Building 212 from 1969 to 1987, then the maximum ambient neutron dose value 

may be adjusted in accordance with ORAUT-PROC-0060, Occupational Onsite Ambient Dose

Reconstruction for DOE Sites (ORAUT 2006). To be favorable to claimants, a neutron energy range 

of 0.1 to 2.0 MeV should be chosen along with an International Commission on Radiological 
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 28 of 43

Figure 4-6. Main site neutron dosimeter locations (Griggs, Myers, and Buddemeier 1984).

Protection Publication 60 weighting factor correction of 1.91 (ICRP 1991). The value for 1972 should 

be used for 1969 to 1971. If the worker was near Building 212 from 1966 to 1969, the neutron-tophoton ratio can be used. If photon dosimetry data is not available for this period, then the value for 

1972 can be used for 1966 to 1969.

 Table 4-10. Main site external 

gamma radiation (mrem/yr) [7].a

Year

Average, 

8,760 hr/yr

Maximum, 

8,760 hr/yr

1952 89b 317c

1953 89b 317c

1954 89b 317c

1955 89b 317c

1956 89b 317c

1957 89b 317c

1958 89b 317c
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 29 of 43

Year

Average, 

8,760 hr/yr

Maximum, 

8,760 hr/yr

1959 50 317c

1960 89b 317c

1961 88 317c

1962 44 317c

1963 50 317c

1964 50 317c

1965 50 317c

1966 89b 317c

1967 20 317c

1968 20 163

1969 34 83

1970 15 105

1971 77 257

1972 89 122

1973 80 96

1974 74 137

1975 88 317

1976 84 293

1977 79 227

1978 64 93

1979 61 96

1980 63 81

1981 54 61

1982 49 57

1983 51 73

1984 48 53

1985 59 80

1986 61 70

1987 65 72

1988 63 70

1989 63 75

1990 65 72

1991 65 78

1992 66 76

1993 65 74

1994 72 80

1995 56 61

1996 55 60

1997 60 64

1998 60 66

1999 58 63

2000 57 65

2001 56 100

2002 65 65

2003 56 62

2004 57 64

2005 58 65

a. The results in this table reflect data 

 provided by LLNL for an exposure period

 of 8,760 hr/yr. The dose values based on

 8760 hr/yr per year should be adjusted in 

 accordance with ORAUT 2006.

b. External dose for these years was not 

 available. The maximum average 

 external
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 30 of 43

 dose was reported in 1972. Therefore,

 the 1972 dose values are used for the 

 noted years.

c. External dose for these years was not 

 available. The maximum external dose

 was reported in 1975. Therefore, the 1975

 dose values are used for the years noted.

 Table 4-11. Site 300 external 

 gamma radiation (mrem/yr) [8].a

Year

Average, 

8,760 hr/yr

Maximum, 

8,760 hr/yr

1955 88b 99b

1956 88b 99b

1957 88b 99b

1958 88b 99b

1959 88b 99b

1960 88b 99b

1961 88b 99b

1962 88b 99b

1963 88b 99b

1964 88b 99b

1965 88b 99b

1966 88b 99b

1967 88b 99b

1968 88b 99b

1969 88b 99b

1970 88b 99b

1971 88b 99b

1972 88b 99b

1973 88b 99b

1974 88b 99b

1975 88b 99b

1976 88b 99b

1977 88b 99b

1978 88b 99b

1979 88b 99b

1980 88b 99b

1981 88b 99b

1982 88b 99b

1983 88b 99b

1984 88b 99b

1985 88b 99b

1986 88b 99b

1987 88b 99b

1988d 88b 99b

1989 77 89

1990 78 94

1991 78 93

1992 77 87

1993 76 86

1994 88 99

1995 64 72

1996 66 72

1997 72 77

1998 72 78

1999 71 80
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 31 of 43

Year

Average, 

8,760 hr/yr

Maximum, 

8,760 hr/yr

2000 64 70

2001 64 66

2002 67 76

2003 67 74

2004 69 77

2005 69 83

a. The results in this table reflect data 

 provided by LLNL for an exposure period 

 of 8,760 hr/yr. The dose values based on

 8760 hr/yr should be adjusted in 

 accordance with ORAUT 2006.

b. External dose for these years was not 

 available. The average and maximum 

 external doses were reported in 1994. 

 Therefore, the 1994 dose values are used 

 for the years noted. 

c. Site average (millirem per year) is based 

 on the maximum value from later years

 where monitoring was performed. 

 Therefore, no maximum or uncertainty is 

 provided.

d. Environmental radiation monitoring at

 Site 300 began on July 1, 1988. 

 Table 4-12. Main site external 

 neutron radiation (mrem/yr) [9].a

Year

Maximum, 

8,760 hr/yr

Maximum,

8,760 hr/yr)

1972 250 57

1973 250 57

1974 370 84

1975 700 160

1976 600 137

1977 550 126

1978 137 31

1979 80 18

1980 85 19

1981 33 8

1982 36 8

1983 113 26

1984 45 10

1985 7 2

1986 30 7

1987 9 2

1988 5 1

1989 6 1

1990 6 1

1991 7 2

1992 7 2

1993 7 2

a. The results in this table reflect data

 provided by LLNL for an exposure 

 period of 8,760 hr/yr. The dose values

 based on 8760 hr/yr should be adjusted 

 in accordance with ORAUT 2006.
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 32 of 43

4.5 ATTRIBUTIONS AND ANNOTATIONS

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

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

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

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

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

for review on the Project’s Site Research Database (SRDB).

[1] Szalinski, Paul A. Integrated Environmental Management. Health Physicist. April 2007. 

The surrogate radionuclides were derived from LLNL NESHAPs reports (Gallegos et al. 1998,

2000). The following is a quote from the reports for the Livermore site: “In addition, isotopic 

analyses of mixtures of radionuclides are not always available, and radionuclide inventories 

are stated as ‘gross alpha,’ ‘gross beta,’ ‘gross gamma,’ or ’mixed fission products’.” In these 

cases, 239Pu was used as the surrogate for gross alpha, 137Cs was used as the surrogate for 

gross gamma, and 90Sr was used as the surrogate for gross beta and mixed fission products 

to provide conservative dose estimates.”

[2] Szalinski, Paul A. Integrated Environmental Management. Health Physicist. April 2007. 

The surrogate radionuclides for gross alpha at Site 300 were derived from LLNL NESHAPs 

reports (Gallegos et al. 1998, 2000). For Site 300 the following applies for gross alpha 

analyses: “During Site 300 explosives experiments, the device containing depleted uranium is 

placed on an open air firing table and detonated.”

[3] Szalinski, Paul A. Integrated Environmental Management. Health Physicist. April 2007. 

The surrogate radionuclide for gross beta at Site 300 is 234Th. Only tritium and depleted 

uranium were reported as released into the environment at Site 300 (Gallegos et al. 1998, 

2000). 234Th is a progeny of the 238

U.

[4] Szalinski, Paul A. Integrated Environmental Management. Health Physicist. March 2007. 

A GSD of 3 is used in ORAUT-OTIB-0018, Internal Dose Overestimates for Facilities with Air 

Sampling Programs (ORAUT 2005). 

[5] Szalinski, Paul A. Integrated Environmental Management. Health Physicist. March 2007. 

200 TBq was arrived at as an annual average by assuming that the 4 PBq that was not 

attributed to the two major release events was released over a 20-year period rather than the 

actual 40-year period. This results in an annual release that is overestimated and favorable to 

the claimant.

[6] Szalinski, Paul A. Integrated Environmental Management. Health Physicist. March 2007. 

The assumption that all the tritium is HTO is favorable to the claimant because, according to 

ICRP Publication 68, HTO has a much higher dose coefficient than tritium gas (ICRP 1995). 

[7] Szalinski, Paul A. Integrated Environmental Management. Health Physicist. April 2007. 

The external gamma radiation data are directly from site annual environmental reports except 

where indicated by footnote. The site maximum and average values from the reports are 

listed in the table.

[8] Szalinski, Paul A. Integrated Environmental Management. Health Physicist. April 2007. 

The external gamma radiation data are directly from site annual environmental reports except 

where indicated by footnote. The site maximum and average values from the reports are 

listed in the table.
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 33 of 43

[9] Smith, Matthew H. Dade Moeller & Associates. Sr. Health Physicist. November 2009. 

The external neutron radiation data are directly from site annual environmental reports. The 

site maximum and average values from the reports are listed in the table.
Document No. ORAUT-TKBS-0035-4 Revision No. 01 Effective Date: 03/16/2010 Page 34 of 43

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1959

and 

1960

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R. Brandstetter, S. L. Brigdon, R. A. Brown, E. Christofferson, K. J. Folks, G. M. 

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2003 Sanchez, L., P. E. Althouse, N. A. Bertoldo, R. G. Blake, S, L. Brigdon, R, A. Brown, E. 

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GLOSSARY

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2. Alpha particles have low penetrating power and a short range (a few 

centimeters in air). Outside the body, the most energetic alpha particle generally fails to 

penetrate the dead layers of cells covering the skin or a sheet of paper. Alpha particles 

emitted by radionuclides inside the body are a more significant health risk. 

background radiation

Radiation from cosmic sources, naturally occurring radioactive materials including naturally 

occurring radon, and global fallout from the testing of nuclear explosives. Background 

radiation does not include radiation from source, byproduct, or Special Nuclear Materials 

regulated by the U.S. Nuclear Regulatory Commission. The average individual exposure from 

background radiation is about 360 millirem per year.

becquerel (Bq)

International System unit of radioactivity equal to 1 disintegration per second; 1 curie equals 

37 billion (3.7 × 1010) Bq.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron. Most of the direct fission products are (negative) beta emitters. 

Exposure to large amounts of beta radiation from external sources can cause skin burns 

(erythema), and beta emitters can be harmful inside the body. Thin sheets of metal or plastic 

can stop beta particles.

deep dose equivalent

Dose equivalent in units of rem or sievert for a 1-centimeter depth in tissue (1,000 milligrams 

per square centimeter).

dosimetry

Measurement and calculation of internal and external radiation doses.

radioactivity

Disintegration of certain elements (e.g., radium, actinium, uranium, and thorium) accompanied 

by the emission of alpha, beta, gamma, and/or neutron radiation from unstable nuclei.

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The average American receives 360 millirem a year from background radiation. The 

sievert is the International System unit; 1 rem equals 0.01 sievert. The word derives from 

roentgen equivalent in man; rem is also the plural.

thermoluminescent dosimeter (TLD) 

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated by radiation, release the stored energy as light. The measurement of this 

light provides a measurement of absorbed dose.