ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 66

Document Title:

Site Profile for the Lawrence Berkeley National 

Laboratory

Document Number: ORAUT-TKBS-0049

Revision: 01 

Effective Date: 04/02/2007

Type of Document TBD

Supersedes: Revision 00

Subject Expert(s): Paul A. Szalinski, Baynard Turpin, James E. Turner, James East, 

 Elizabeth A. Langille, and Bill R. Thomas

Site Expert(s): 

Approval: Signature on File Approval Date: 03/27/2007

Paul A. Szalinski, Document Owner

Concurrence: Signature on File Concurrence Date: 03/27/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 03/27/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 03/27/2007

Kate Kimpan, Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 04/02/2007

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 2 of 66

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

08/07/2006 00 New site profile for the Lawrence Berkeley National Laboratory. First 

approved issue. Training required: As determined by the Task 

Manager. Initiated by Paul A. Szalinski.

04/02/2007 01 Approved Revision 01 initiated to include Task 5 and PID review 

comments which were omitted from Rev 00. Constitutes a total 

rewrite of document. Incorporates internal formal review comments 

and additional NIOSH formal review comments. Attributions and 

Annotations added. Incorporates internal formal review comments 

regarding the Attributions and Annotation section only. This revision 

results in an increase in assigned dose and a PER is required. 

Training required: As determined by the Task Manager. Initiated by 

Paul A. Szalinski.
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 3 of 66

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

1.0 Introduction ..............................................................................................................................7

1.1 Purpose........................................................................................................................8

1.2 Scope...........................................................................................................................8

2.0 Site Description........................................................................................................................8

3.0 Occupational Medical Dose.................................................................................................... 18

4.0 Occupational Environmental Dose ......................................................................................... 20

5.0 Occupational Internal Dose.................................................................................................... 24

6.0 Occupational External Dose ................................................................................................... 39

7.0 Attributions and Annotations .................................................................................................. 43

References ........................................................................................................................................50

Glossary ............................................................................................................................................59
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 4 of 66

LIST OF TABLES

TABLE TITLE

2-1 Area information and parameters.............................................................................................

PAGE

9

2-2 Magnitude of activities by area............................................................................................... 11

3-1 Parameters and organ doses for 14- by 17-in. PA chest radiography..................................... 19

3-2 Maximum organ doses by year for 14- by 17-in. PA chest radiography.................................. 19

4-1 External gamma and neutron radiation................................................................................... 21

4-2 Maximum site-wide annual median intakes via inhalation....................................................... 23

5-1 Internal dose control program (in vitro) ................................................................................... 26

5-2 Sample type (in vitro) ............................................................................................................. 27

5-3 Internal dose control program (in vivo) ................................................................................... 27

5-4 In vitro detection limits............................................................................................................ 28

5-5 In vivo detection limits ............................................................................................................ 31

5-6 Units codes and description of units....................................................................................... 32

5-7 1949 Chalk River tolerances, maximum permissible doses in inspired air.............................. 33

5-8 Radionuclides and fraction activity by facility.......................................................................... 34

6-1 LBNL dosimeter type, period of use, exchange frequency, MDL, and 

potential annual missed dose................................................................................................. 40

6-2 Selection of beta and photon radiation energies and percentages ......................................... 41

6-3 Selection of neutron radiation energies and percentages....................................................... 41

6-4 Recorded dose practices........................................................................................................ 41

6-5 Interpretation of reported data................................................................................................ 42

6-6 Adjustments to recorded dose................................................................................................ 42

6-7 Bias and uncertainty...............................................................................................................43
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 5 of 66

ACRONYMS AND ABBREVIATIONS

ALARA as low as reasonably achievable

AMAD activity median aerodynamic diameter

ANL-E Argonne National Laboratory-East

Bq becquerel

Ci curie

cm centimeter

cpm counts per minute

d day

DOE U.S. Department of Energy

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ESE entrance skin exposure

F fast (solubility rate)

ft foot

g gram

HILAC heavy-ion linear accelerator

hr hour

HVL half-value layer

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis (computer program)

in. inch

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

kV kilovolt

kVp applied kilovoltage; peak kilovoltage

L liter

LANL Los Alamos National Laboratory

LBNL Lawrence Berkeley National Laboratory

LLNL Lawrence Livermore National Laboratory

LRL Lawrence Radiation Laboratory

LRL-B Lawrence Radiation Laboratory Berkeley

m meter

M moderate (solubility rate)

mA milliampere

MDA minimum detectable activity

MDL minimum detection limit

MeV megaelectron-volt, 1 million electron-volts

MFP mixed fission products
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 6 of 66

min minute

mL milliliter

mm millimeter

mo month

MPC Maximum Permissible Concentration

mR milliroentgen

mrem millirem

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH-Office of Compensation Analysis and Support Claims Tracking System 

NTA nuclear track emulsion, type A

NTLF National Tritium Labeling Facility

ORAU Oak Ridge Associated Universities

PA posterior-anterior

pCi picocurie

PHA pulse height analysis

POC probability of causation

S slow (solubility rate)

s second

SRDB Ref ID Site Research Database Reference Identification Number

TLD thermoluminescent dosimeter

U.S.C. United States Code

WB whole-body

WBC whole-body count

wk week

yr year

μg microgram

μm micrometer

μCi microcurie

§ section or sections
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 7 of 66

1.0

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.

INTRODUCTION

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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

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

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

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

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

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

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

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

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

external dosimetry monitoring results are considered valid for use in dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

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

derived:

] guidelines established under 

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

guidelines (nor the dose reconstruction regulation) define “performance of duty” for DOE employees 

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

• Radiation from naturally occurring radon present in conventional structures

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

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 8 of 66

1.1 PURPOSE

This site profile provides technical basis information to be used to evaluate the total occupational 

radiation dose that can reasonably be associated with a worker’s radiation exposure at Lawrence 

Berkeley National Laboratory (LBNL). This dose results from exposure to external and internal 

radiation sources in LBNL facilities, to X-ray examinations performed for medical screening, and to 

onsite environmental releases. This site profile includes methods for estimating doses that could 

have occurred while an employee was not monitored or was inadequately monitored as well as doses 

that were missed due to analytical detection limits or incomplete or missing monitoring records (i.e., 

missed dose). 

This site profile can be a tool when performing dose reconstructions for LBNL workers. The 

Integrated Modules for Bioassay Analysis (IMBA) computer program is a tool useful for internal dose 

calculations. Information on measurement uncertainties is an integral component of the NIOSH 

approach. This document describes how to evaluate uncertainty associated with LBNL exposure and 

dosimetry records

1.2 SCOPE

This site profile consists of a site description (Section 2.0) and discussions of occupational medical 

dose, occupational environmental dose, occupational internal dose, and occupational external dose 

(Sections 3.0 to 6.0).

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 7.0.

2.0

This section describes major LBNL facilities and operations. LBNL was founded in 1931 on the 

University of California, Berkeley campus. It was originally known as the University of California 

Radiation Laboratory. In July 1945, LBNL moved from the Berkeley campus to a 134-acre site in the 

hills to the east [1]. LBNL began defense work in 1941 when the National Defense Research 

Committee appointed Ernest O. Lawrence to study the potential military uses of 235U (Maroncelli and 

Karpin 2002). Assignment of dose should begin in 1942 when the Manhattan Engineer District was 

founded. 

SITE DESCRIPTION

LBNL is alternatively referred to in reference documents as Berkeley Lab, Lawrence Radiation 

Laboratory (LRL), and Lawrence Radiation Laboratory, Berkeley (LRL-B). This document uses LBNL 

for consistency.

Table 2-1 provides a general description of LBNL buildings and lists the radionuclides of concern [2]. 

From the 1950s to the present, the laboratory has maintained its status as a major international center 

for physics research and has diversified its research program into almost every realm of scientific 

investigation. Along with its historical specialty of accelerator research and nuclear physics, the 

laboratory currently maintains divisions that investigate astrophysics, nuclear fusion, earth sciences, 

genomics, health physics, computer science, materials science, environmental science, and other 

areas [3]. In addition, the laboratory is the site of a number of National User Facilities, including the 

Advanced Light Source, the National Center for Electron Microscopy, the National Energy Research 

Scientific Computing Center, the Energy Sciences Network, and the future Molecular Foundry [4]. 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 9 of 66

Table 2-2 lists the quantities of the radionuclides that workers could have encountered by area [5]. 

This list is not intended as a complete radiological history, but rather as a discussion to familiarize 

dose reconstructors with the variety of radionuclides that have been present at LBNL.

Table 2-1. Area information and parameters.

a

Building(s) Description Period Radionuclidesb

1 (on campus) Donner Laboratory 1931 H-3, C-14, P-32, P-33, Sr-90, I-125, Th-232, U-238

2 (on campus) Advanced Materials 

Laboratory and Center 

for X-ray Optics (formerly 

the Crocker Laboratory)

1931 N-13, O-15

3 Laboratory of Chemical 

Biodynamics (Calvin 

Laboratory)

H-3, C-14, P-32, S-35, P-33

4 and 5 Magnetic Fusion Energy H-3, C-14, P-32, S-35

6, 9, 10, and 

80

Advanced Light Source 

(formerly the 184-in. 

cyclotron complex)

1940 to 

present

Be-7, Co-57, Co-58, Co-60, Ni-63, Fe-55, Fe-59, Mn-54, 

Na-2,2 Zn-65, Am-241, Am-243, Cm-246, Cm-248, 

Np-237, Pu-238, Pu-239, Pu-240, U-235, Tc-99, Pu-242, 

Pu-241, N-13, O-15, Na-24, Ar-41, Ni-57, Th-232

8 (on campus) Hearst Laboratory Am-241, U

10 Cell & Molecular Biology 

Research & Photography

1986 H-3, C-14, P-32, S-35, Cr-51

14 Accelerator and Fusion 

Research and Earth 

Sciences

H-3 

16 Magnetic Fusion Energy 

Laboratory

U-234, U-235, U-238

19 (on 

campus)

LeConte Hall Fe-55, Fe-59, Co-60, Au-198, Au-199, Hg-203, Ga-67, 

In-112, Ba-127, Cs-127, Ba-131, Cs-131, Cs-137, Sr-90

22 (on 

campus)

Latimer Hall C-14

26/76 Medical Services and 

Bioassay, Radiation and 

Analytical Measurements 

Laboratory

I-125, I-129, I-131, U-238, C-11, C-14, Be-7, F-18, P-32, 

S-35, U-232, Sr-90 Ac-228, Ag-108, Ba-133, Ce-144, 

Cm-243, Am-241, Am-243, Br-76, Br-77, Cm-244, Bk-249, 

Cf-249, Cf-250, Cm-245, Cm-246

31 (on 

campus)

Hesse Court Depleted U

36 (on 

campus)

2232 Piedmont Sr-90, Am-241, Na-24, Cl-36, K-42

38 (on 

campus)

Lewis Hall Co-57, Fe-59, Co-60, Sn-119, Gd-153

50B Physics Ag-106m, Au-194, Co-56, Co-57, Co-58, Co-60, Mn-52, 

Mn-54, Na-22, Ni-57, Re-184m, Sc-48, Se-75, Ta-182, 

V-48, Zn-65m Be-7, Ag-105

51 Contains the Bevatron 

and Bevalac

Be-7, Co-57, Co-58, Co-60, Fe-55, Fe-59, Mn-54, Na-22, 

Ni-63, Zn-65, Na-24, Ni-57, Ni-57, Cf-252, U, Am-241, 

Am-243, Ra-226, Bk-249

52 Magnetic Fusion Energy 

Laboratory

H-3, C-14, P-32, S-35, Am-241, Cf-250, Cm-243, Cm-248, 

Np-237, Pu-239, Pu-244, Th-229, Th-230, Th-232, U-232, 

U-235, U-236, U-238, Pu-242

53 SuperHILAC 

Development

55/56 Research 

Medicine/Radiation 

Biophysics, Biomedical 

Isotope Facility

1964 H-3, C-14, F-18, P-32, I-131, Sr-90, Xe-127, Nb-95, 

Ru-103, Gd-153, Tl-201, I-123, I-125, Sn-113, Tc-99m, 

Co-57, Cr-51, Nb-95, O-15, Cu-64, Ce-141, N-13, Sr-85, 

Sc-46, C-11, Co-55,

56 Biomedical Isotope 

Facility

C-11, Co-55, Co-57, F-18, N-13, O-15, O-14, F-17 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 10 of 66

Building(s) Description Period Radionuclidesa

57 (on 

campus)

Donner Pavilion P-32, Sr-82, Sr-85, Sr-90, I-125, I-131, In-111

58 and 58A Accelerator Research 

and Development

60 High Bay Laboratory

62 Materials and Chemical 

Sciences

U-238

63 and 64 Accelerator and Fusion 

Research

P-32

70 and 70A Nuclear, Materials, 

Chemical, Life, and Earth 

Sciences

H-3, C-14, P-32, Fe-59, I-125, Sr-90, Th-232, Tc-99, 

Ce-141, Ce-144, Pa-233, Sm-153, Te-153m, Yb-169, 

Zn-65, Sc-46, Na-22, Na-24, S-35, Ca-45, Ca-49, Mn-54, 

Fe-59, Y-90, Rh-101, Ru-106, Sb-124, Ce-141, Tm-170, 

Cs-134, Eu-152, Am-243, Bi-207, Sc-49, Co-60, Zr-88, 

Rb-86, Zr-95, Sb-122, Ta-182, Tl-204, Th-229, Yb-175, 

Ho-166m, Hf-175, Ac-227, Pa-233, Am-241, Cm-244, 

U-233, U-234, U-235, U-238, Pu-238, Pu-239, Ra-226, 

Np-237, Cf-249, Cf-252, Tb-161, Cm-248

71 HILAC 1957 to 

December 

23, 1992

Sc-46, Sc-49, Sr-90, Zr-88, Y-90, Zr-95, Tc-99, Sb-124,

Tb-161, Tm-170, Yb-175, Ta-182, Tl-204, Th-229, Th-232, 

U-233, U-234, U-235, U-238

72 Health Physics 1961 P-32, Na-24, Ti-44, Cr-51 Mn-54, Co-58, Fe-59, Co-60, 

Cu-64, Zn-65, Br-82, Nb-95, Zr-95, Zr-97, Cd-113, Sb-122, 

Sb-124, Ba-133, Xe-133, Eu-152, Ta-182, Hg-194, Au-198, 

Bi-207, Ac-227, Th-228, Th-229, Pa-231, Th-232, U-238

73 Atmospheric Aerosol 

Research

H-3 

74 Research 

Medicine/Radiation 

Biophysics, Cell and 

Molecular Biology

H-3, C-11, N-13, C-14, O-15, F-18, P-32, S-35, Sc-46, 

Fe-59, Cu-64, Sr-85, Nb-95, Tc-99m, Ru-103, I-123, I-125, 

I-131, Tl-201, Pb-210, Th-232, U-233, U-234/235, U-238, 

Pu-238, Am-241

75 Radioisotope Service a 

National Tritium Facility 

H-3, Sr-90, Th-232, Act. Products, C-11, C-14, F-18, I-125, 

I-129, I-131, P-32, S-35, Sr-90, U-232, Kr-81, Kr-85

75A Compact, Processing 

and Storage Facility

Th-232, Sr-90

76 C-11, C-14, F-18, I-125, I-129, I-131, P-32, S-35, Sr-90, 

U-232, Am-241

84 H-3, C-14, P-32, S-35

85 Hazardous Waste 

Handling Facility

H-3, C-14, Sr-90, I-125, Th-232

88 88-in. Cyclotron Be-7, C-11, N-13, C-14, O-14, O-15, F-17, F-18, Na-21, 

Na-22, Na-24, P-32, S-35, Sc-46, Cr-51, Mn-54, Fe-55, 

Co-56, Co-57, Ni-57, Co-58, Fe-59, Co-60, Cu-60, Zn-62, 

Zn-63, Ni-63, Zn-65, Ge-71, Se-75, Br-76, Br-77, Kr-76, 

Kr-77, Kr-79, Y-88, Sr-90, Zr-95, Zr-97, U-238, Sc-93, 

Ac-227, Am-241, Au-198, Cs-134, Cs-137, Gd-148, 

Eu-152, Eu-154, Pb-212, Np-237, Np-239, Th-229, 

Th-232, Pa-233, Pu-238, Pu-239, Sb-124, Pa-231, Cf-249, 

Cf-252

934 Molecular and Cell 

Biology

H-3, P-32, S-35, C-14, I-125

977 Life Sciences and 

Physical Biosciences

H-3, P-32, P-33, S-35, C-14, Cd-109

a. Blank values indicate not applicable or data not available.

b. Particle size assumed as the default value of 5-μm activity median aerodynamic diameter (AMAD) because no site-specific data 

have been found.
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 11 of 66

Table 2-2. Magnitude of activities by area. 

Building Radionuclides Activity (Ci)a

1, Room 159 K-42 0.001

F2-65 0.001

Zn-65 0.001

Sr-85 0.001

I-131 0.001

1, Room 212 Fe-59 0.001

1, Room 213 C-14 Trace

1, Room 216 C-14 0.005

P-32 0.03

Cr-51 0.01

Fe-55 0.01

Fe-59 0.01

Sr-90 0.02

Y-90 0.1

I-131 0.02

Hg-203 0.01

1, Room 227 P-32 0.05

I-131 0.25

1, Room 230 Ca-47 0.001

Zn-65 0.001

I-131 0.001

1, Room 261 P-32 0.5

I-131 0.001

1, Room 264 F-18 0.01

Fe-52 0.1

Ge-68 0.001

As-71 0.1

I-123 0.001

1, Room 271 Na-24 0.001

P-32 0.5

1, Room 308 C-14 0.005

1, Room 322 P-32 (b)

1, Room 326 P-32, P-33 (b)

1, Room 330 H-3 Tracers

1, Room 361 P-32 (b)

1, Room 364 F-18 1E-6

Fe-52 1E-6

Fe-59 1E-6

Ga-68 1E-6

As-71 1E-6

I-131 1E-6

Hg-203 1E-6

1, Room 366 P-32, P-33 (b)

1, Room 373 H-3, C-14, P-32, P33, S-35, I-125 (b)

1, Room 471 P-32 0.005

2, Rooms 101 and 

101A

Ca-45 Tracers

Sr-85 Tracers

Sr-90 Tracers

Am-241 Tracers

2, Room 102 Na-22 0.005

Ca-45 0.001

Sr-85 0.001

Sr-90 0.002

Eu-152/Eu-154 0.0015

Ac-227 0.003

Am-241 0.00015
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 12 of 66

Building Radionuclides Activity (Ci)a

2, Room 103 Ca-45 Tracers

2, Room 104 Sr-85 Tracers

2, Room 108 Sr-90 Tracers

2, Room 116 Am-241 Tracers

3, Room 120 C-14 0.01

3, 2nd floor lab H-3 5E-6

C-14 5E-6

P-32 5E-6

3, Room 214 H-3 30

C-14 0.05

Varied Isotope storage

3, Room 250 C-14, P-32, P-33, S-35 (a)

3, 3rd floor lab H-3 5E-6

C-14 5E-6

P-32 5E-6

3, Room 318 H-3 Varied

C-14 0.03

P-32 Varied

3, Room 322 Varied Calibration standards

3, Room 326 C-14 30

4 Cs-137 0.2 (July 1963)c

5 Co-60 3,100 (1955)c

6 Am-241, Am-243, Cm-248, Eu-152, Eu-154, Np-237, Pu-239, Sr-90, 

Th-232, U-233, U(natural), U-235, U-238, Pu-241, Pu-242, Tc-99 

(b)

8, Room 360 Am-241 0.0066

U 1 kg, various compounds

10 Cs-137 1,050 (1968)c

16, Room 101 U-234, U-235, U-238 (b)

19 (LeConte Hall) Co-60 0.02

19, Room 79 Varied 0.01, experiment residues

Pb-197/Pb-209 1.5 (cyclotron target)

19, Room 81 Varied 0.01, experiment residues

19, Room 86 Fe-55/Fe-59 1 (reactor capsule)

Ga-67 1.5 (cyclotron target)

In-112 2 (cyclotron target)

Ba-127/Cs-127 1.5 (cyclotron target)

Ba-131/Cs-131 24 (reactor capsule)

Au-198 1.5 (cyclotron target)

Au-199 1.5 (cyclotron target)

Hg-203 0.03

19, Room 405 Sr-90 0.001

Cs-137 0.001

22, Latimer Hall 

Room 710

C-14 Tracers

26, Rooms 6, 24, 

30, 31, and 32

Alpha, Am-241, beta-gamma, C-11, C-14, F-18, H-3, I-125, I-129, 

I-131, P-32, S-35, U-232, Sr-90

(b) 

31, Hesse Court Depleted U 7 kg (gaseous conversion 

of UCl4 to UCl6)

36, 2232 Piedmont

basement

Sr-90

Am-241

0.001

0.001

36, 2232 Piedmont

kitchen

Na-24

Cl-36

K-42

0.001

0.001

0.001

38, Lewis Hall Co-57/Co-60 0.02

Fe-59 0.025

Sn-119 0.0005

Gd-153 0.001
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 13 of 66

Building Radionuclides Activity (Ci)a

50 Ag-106m, Au-194, Co-56, Co-57, Co-58, Co-60, Mn-52, Mn-54, Na-22, 

Ni-57, Re-184m, Sc-48, Se-75, Ta-182, v-48, Zn-65m Be-7, Ag-105

(b)

50B, Room 6209 Ag-106m, Au-194, Co-56, Co-57, Co-58, Co-60, Mn-52, Mn-54, Na-22, 

Ni-57, Re-184m, Sc-48, Se-75, Ta-182, V-48, Zn-65, Be-7, Sc-46, 

Ag-105

(b)

51 Be-7, Co-57, Co-58, Co-60, Fe-55, Fe-59, Mn-54, Na-22, Ni-63, Zn-65, 

Na-24, Ni-57, Ni-57, Cf-252, U, Am-241, Am-243, Ra-226, Bk-249

(b)

52, Rooms 109 

and 111

Act Products, Am-241, Cf-250, Cm-243, Cm-248, Np-237, Pu-239, 

Pu-244, Th-229, Th-230, Th-232, U-232, U-235, U-236, U-238, Pu-242

(b)

55 Sr-90 30 (spill - 1964)c

55, Room 116 H-3, C-14, Co-57, Cr-51, F-18, Gd-153, H-3, I-125, I-131, Nb-95, 

Ru-103, Sn-113, Tc-99m, Tl-201 

(b)

55, Room 118 C-14, Co-57, Cr-51, F-18, Gd-153, H-3, I-125, I-131, Nb-95, Ru-103, 

Sn-113, Tc-99m, Tl-201 

(b)

55, Room 120 H-3, C-11, C-14, Co-55, Co-57, Ce-141, Cu-64, F-18, I-123, I-125, 

I-131, N-13, Nb-95, O-15, Ru-103, Sr-85, Tc-99m, Gd-153, Tl-201, 

Sc-46 

(b) 

55, Room 122 H-3, C-11, C-14, F-18, Ce-141, Cu-64, F-18, I-123, I-125, I-131, N-13, 

Nb-95, O-15, Ru-103, Sr-85, Tc-99m, Tl-201, Sc-46

(b)

55, Room 126 H-3, C-11, C-14, F-18, Ce-141, Cu-64, F-18, I-123, I-125, I-131, N-13, 

Nb-95, O-15, Ru-103, Sr-85, Tc-99m, Tl-201, Sc-46

(b)

55, Room 128 H-3, C-11, C-14, F-18, Ce-141, Cu-64, I-123, I-125, I-131, N-13, Nb-95, 

O-15, Ru-103, Sr-85, Tc-99m, Tl-201, Sc-46, Co-55, Co-57, I-123, 

I-125, I-131, Tc-99m, I-125, Gd-153, Sn-113

(b)

55, Room 134 C-11, C-14, Co-55, Co-57, F-18, H-3, I-123, I-125, I-131, Tc-99m, N-13, 

O-15, O-14, F-17, Ce-141, Cu-64, Nb-95, Ru-103, Sr-85, Tl-201, Sc-46, 

Cr-51, Gd-153, H-3, Sn-113, Tl-201

(b)

55, Room 136 C-14, F-18, H-3, I-125 (b)

55, Room 139 C-11, C-14, Cr-51, Co-55, Co-57, F-18, H-3, I-123, I-125, I-131, Tc99m, Ce-141, Cu-64, N-13, Nb-95, O-15, Ru-103, Sr-85, Tl-201, Sc-46(b) 

55, Room 139A C-11, C-14, Co-55, Co-57, F-18, H-3, I-123, I-125, I-131, Tc-99m, 

Ce-141, Cu-64, N-13, Nb-95, O-15, Ru-103, Sr-85, Tl-201, Sc-46

(b) 

55, Room 151 C-11, C-14, Co-55, Co-57, F-18, H-3, I-123, I-125, I-131, Tc-99m, N-13, 

O-15, O-14, F-17, Ce-141, Cu-64, Nb-95, Ru-103, Sr-85, Tl-201, Sc-46, 

Cr-51, Gd-153, H-3, Sn-113, Tl-201

(b)

55, Room 200 C-11, Ce-141, Cu-64, F-18, I-123, I-125, I-131, N-13, Nb-95, O-15, 

Ru-103, Sr-85, Tc-99m, Tl-201, Sc-46

(b)

56, Room 100 C-11, Co-55, Co-57, F-18, N-13, O-15, O-14, F-17 (b)

56, Room 101 C-11, Co-55, Co-57, F-18, N-13, O-15, O-14, F-17 (b)

57, Cowell 

Hospital, 3rd floor

Y-90 1

57, Cowell 

Hospital, 4th floor

I-125 0.01

57, Cowell 

Hospital, 4th floor

I-131 0.001

62, Rooms 114 

and 145

U-238

64, Room 234 P-32 (b)

69, Room 150 Various activation products (b)

70 Cm-242/Pu-238

70, Room 103 Ce-141, Ce-144, Fe-59, H-3, P-32, Pa-233, Sm-153, Te-125m, Yb-169, 

Zn-65, Sc-46

(b) 

70, Room 108C Ce-141, Ce-144, Fe-59, H-3, P-32, Pa-233, Sm-153, Te-125m, Yb-169, 

Zn-65, Sc-46

(b)

70, Room 114 U-238 (b)

70, Room 114A U-238 (b)

70, Room 133 Ce-141, Ce-144, Fe-59, H-3, P-32, Pa-233, Sm-153, Te-125m, Yb-169, 

Zn-65, Sc-46

(b) 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 14 of 66

Building Radionuclides Activity (Ci)a

70, Room 147 Am-241, Cf-250, Cm-243, Cm-248, Np-237, Pu-239, Pu-244, Th-229, 

Th-230, Th-232, U-232, U-235, U-236, U-238, Pu-242

(b)

70, Room 147A Am-241, Am-243, Cf-249, Cm-243, Cm-245, Cm-246, Cm-248, Co-60, 

Eu-152, Eu-154, Eu-155, Ni-65, Np-237, Np-239, Pa-233, Pu-238, 

Pu-239, Pu-240, Pu-244, Ra-226, Ru-106, Th-229, Th-230, Th-232, 

U-232, U-232, U-233, U-235, U-238, Tc-99, Pa-231, Pu-242, Cm-244, 

Cs-137, Ni-63, Pu-241, Sr-90, U-236, Bk-249, Cf-250, U-234, Es-254, 

P-32, Na-21, Na-22, Co-57, Cu-60, Na-22, Ni-57, Sc-93, Zn-62, Zn-63, 

Ac-227, Au-198, C-11, Cf-249, Cf-252, Co-56, Co-57, Co-58, Cr-51, 

Cs-134, Cs-137, Fe-55, Fe-59, Mn-54, Sb-124, Se-75, Zn-65, Be-7, 

Sc-46, Pa-231

(b)

70, Room 203 Am-241, Am-243, Bi-207, Cf-249, Cf-252, Cm-248, Er-165, Er-169, 

Er-171, Es-253, Eu-152, Fm-257, Hf-172, Hf-175, Hf-181, Ho-166, 

Ho-166m, Na-22, Nb-95, Np-237, Pu-238, Pu-239, Ra-226, Rh-101, 

Rh-102, Ta-182, Tb-160, Th-229, Tm-170, U-233, U-235, U-238, Y-90, 

Zr-88, Zr-95, Cm-244, Mn-54, Sr-90, Tl-204, Bk-249, Es-254

(b)

70, Room 209 Am-241, Am-243, Bi-207, Cf-249, Cf-252, Cm-248, Er-165, Er-169, 

Er-171, Es-253, Eu-152, Fm-257, Hf-172, Hf-175, Hf-181, Ho-166, 

Ho-166m, Na-22, Nb-95, Np-237, Pu-238, Pu-239, Ra-226, Rh-101, 

Rh-102, Ta-182, Tb-160, Th-229, Tm-170, U-233, U-235, U-238, Y-90, 

Zr-88, Zr-95, Cm-244, Mn-54, Sr-90, Tl-204, Bk-249, Es-254

(b)

70, Room 210 Am-241, Am-243, Bi-207, Cf-249, Cf-252, Cm-248, Er-165, Er-169, 

Er-171, Es-253, Eu-152, Fm-257, Hf-172, Hf-175, Hf-181, Ho-166, 

Ho-166m, Na-22, Nb-95, Np-237, Pu-238, Pu-239, Ra-226, Rh-101, 

Rh-102, Ta-182, Tb-160, Th-229, Tm-170, U-233, U-235, U-238, Y-90, 

Zr-88, Zr-95, Cm-244, Mn-54, Sr-90, Tl-204, Bk-249, Es-254

(b)

70A Co-60 9,800 (July 1965)c

70A, Room 1103 C-11, C-14, H-3, P-32 (b)

70A, Room 1115 C-11, C-14, P-32, P-33 (b)

70A, Room 117A C-11, H-3 (b)

70A, Room 1121A C-11, H-3 (b)

70A, Room 1121B C-11, H-3 (b)

70A, Room 1129 Co-60, Ni-63, Ni-65, Sr-90, Tc-99, Cs-137, Eu-152, Eu-154, Eu-155, 

Am-241, Am-243, Bk-249, Cf-249, Cm-243, Cm-244, Cm-245, Cm-246, 

Cm-248, Cf-249, Cf-250, f-252, Es-254, Np-237, Np-239, Pa-231, 

Pa-233, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Pu-244, Ra-226, 

Ru-106, Th-229, Th-230, Th-232, U-232, U-233, U-234, U-235, U-236, 

U-238

(b)

70A, Room 1145 Co-60, Ni-63, Ni-65, Sr-90, Tc-99, Cs-137, Eu-152, Eu-154, Eu-155, 

Am-241, Am-243, Bk-249, Cf-249, Cf-250, Cf-252, Cm-243, Cm-244, 

Cm-245, Cm-246, Cm-248, Cf-249, Es-254, Np-237, Np-239, Pa-231, 

Pa-233, Pu-238, Pu-239, Pu-240, Pu-241, Pu-242, Pu-244, Ra-226, 

Ru-106, Th-229, Th-230, Th-232, U-232, U-233, U-234, U-235, U-236, 

U-238, U(natural)

(b)

70A, Room 1145A Am-241, Am-243, Cf-249, Cm-243, Cm-245, Cm-246, Cm-248, Co-60, 

Eu-152, Eu-154, Eu-155, Ni-65, Np-237, Np-239, Pa-233, Pu-238, 

Pu-239, Pu-240, Pu-241, Pu-244, Ra-226, Ru-106, Th-229, Th-230, 

Th-232, U-232, U-232, U-233, U-235, U-236, U-238, Tc-99, Pa-231, 

Pu-242, Cm-244, Cs-137, Ni-63, Pu-241, Sr-90, Bk-249, Cf-250, U-234, 

Es-254

(b)

70A, Room 1145B Am-241, Am-243, Cf-249, Cm-243, Cm-245, Cm-246, Cm-248, Co-60, 

Eu-152, Eu-154, Eu-155, Ni-65, Np-237, Np-239, Pa-233, Pu-238, 

Pu-239, Pu-240, Pu-241, Pu-244, Ra-226, Ru-106, Th-229, Th-230, 

Th-232, U-232, U-232, U-233, U-235, U-236, U-238, Tc-99, Pa-231, 

Pu-242, Cm-244, Cs-137, Ni-63, Pu-241, Sr-90, Bk-249, Cf-250, U-234, 

Es-254

(b) 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 15 of 66

Building Radionuclides Activity (Ci)a

70A, Room 1149 Am-241, Am-243, Cf-249, Cm-243, Cm-245, Cm-246, Cm-248, Co-60, 

Eu-152, Eu-154, Eu-155, Ni-65, Np-237, Np-239, Pa-233, Pu-238, 

Pu-239, Pu-240, Pu-241, Pu-244, Ra-226, Ru-106, Th-229, Th-230, 

Th-232, U-232, U-232, U-233, U-235, U-236, U-238, Tc-99, Pa-231, 

Pu-242, Cm-244, Cs-137, Ni-63, Pu-241, Sr-90, Bk-249, Cf-250, U-234, 

Es-254

(b)

70A, Room 1151 Am-241, Am-243, Cf-249, Cm-243, Cm-245, Cm-246, Cm-248, Co-60, 

Eu-152, Eu-154, Eu-155, Ni-65, Np-237, Np-239, Pa-233, Pu-238, 

Pu-239, Pu-240, Pu-241, Pu-244, Ra-226, Ru-106, Th-229, Th-230, 

Th-232, U-232, U-232, U-233, U-235, U-236, U-238, Tc-99, Pa-231, 

Pu-242, Cm-244, Cs-137, Ni-63, Pu-241, Sr-90, Bk-249, Cf-250, U-234, 

Es-254

(b) 

70A, Room 1159A Am-241, Am-243, Cf-249, Cm-243, Cm-245, Cm-246, Cm-248, Co-60, 

Eu-152, Eu-154, Eu-155, Ni-65, Np-237, Np-239, Pa-233, Pu-238, 

Pu-239, Pu-240, Pu-241, Pu-244, Ra-226, Ru-106, Th-229, Th-230, 

Th-232, U-232, U-232, U-233, U-235, U-236, U-238, Tc-99, Pa-231, 

Pu-242, Cm-244, Cs-137, Ni-63, Pu-241, Sr-90, Bk-249, Cf-250, U-234, 

Es-254

(b)

70A, Room 1159B Am-241, Am-243, Cf-249, Cm-243, Cm-245, Cm-246, Cm-248, Co-60, 

Eu-152, Eu-154, Eu-155, Ni-65, Np-237, Np-239, Pa-233, Pu-238, 

Pu-239, Pu-240, Pu-241, Pu-244, Ra-226, Ru-106, Th-229, Th-230, 

Th-232, U-232, U-232, U-233, U-235, U-236, U-238, Tc-99, Pa-231, 

Pu-242, Cm-244, Cs-137, Ni-63, Pu-241, Sr-90, Bk-249, Cf-250, U-234, 

Es-254

(b)

70A, Room 1165 Am-241, Am-243, Cf-249, Cm-243, Cm-245, Cm-246, Cm-248, Co-60, 

Eu-152, Eu-154, Eu-155, Ni-65, Np-237, Np-239, Pa-233, Pu-238, 

Pu-239, Pu-240, Pu-241, Pu-244, Ra-226, Ru-106, Th-229, Th-230, 

Th-232, U-232, U-232, U-233, U-235, U-236, U-238, Tc-99, Pa-231, 

Pu-242, Cm-244, Cs-137, Ni-63, Pu-241, Sr-90, Bk-249, Cf-250, U-234, 

Es-254

(b) 

70A, Room 1165A Am-241, Am-243, Cf-249, Cm-243, Cm-245, Cm-246, Cm-248, Co-60, 

Eu-152, Eu-154, Eu-155, Ni-65, Np-237, Np-239, Pa-233, Pu-238, 

Pu-239, Pu-240, Pu-241, Pu-244, Ra-226, Ru-106, Th-229, Th-230, 

Th-232, U-232, U-232, U-233, U-235, U-236, U-238, Tc-99, Pa-231, 

Pu-242, Cm-244, Cs-137, Ni-63, Pu-241, Sr-90, Bk-249, Cf-250, U-234, 

Es-254

(b)

70A, Room 2211 Am-241, Am-243, Cf-249, Cm-243, Cm-245, Cm-246, Cm-248, Co-60, 

Eu-152, Eu-154, Eu-155, Ni-65, Np-237, Np-239, Pa-233, Pu-238, 

Pu-239, Pu-240, Pu-241, Pu-244, Ra-226, Ru-106, Th-229, Th-230, 

Th-232, U-232, U-232, U-233, U-235, U-236, U-238, Tc-99, Pa-231, 

Pu-242, Cm-244, Cs-137, Ni-63, Pu-241, Sr-90, Bk-249, Cf-250, U-234, 

Es-254

(b) 

70A, Room 2215 Am-241, Am-243, Cf-249, Cm-243, Cm-245, Cm-246, Cm-248, Co-60, 

Eu-152, Eu-154, Eu-155, Ni-65, Np-237, Np-239, Pa-233, Pu-238, 

Pu-239, Pu-240, Pu-241, Pu-244, Ra-226, Ru-106, Th-229, Th-230, 

Th-232, U-232, U-232, U-233, U-235, U-236, U-238, Tc-99, Pa-231, 

Pu-242, Cm-244, Cs-137, Ni-63, Pu-241, Sr-90, Bk-249, Cf-250, U-234, 

Es-254

(b)

70A, Room 2217 Am-241, Am-243, Cf-249, Cm-243, Cm-245, Cm-246, Cm-248, Co-60, 

Eu-152, Eu-154, Eu-155, Ni-65, Np-237, Np-239, Pa-233, Pu-238, 

Pu-239, Pu-240, Pu-241, Pu-244, Ra-226, Ru-106, Th-229, Th-230, 

Th-232, U-232, U-232, U-233, U-235, U-236, U-238, Tc-99, Pa-231, 

Pu-242, Cm-244, Cs-137, Ni-63, Pu-241, Sr-90, Bk-249, Cf-250, U-234, 

Es-254

(b)
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 16 of 66

Building Radionuclides Activity (Ci)a

70A, Room 2217B Am-241, Am-243, Cf-249, Cm-243, Cm-245, Cm-246, Cm-248, Co-60, 

Eu-152, Eu-154, Eu-155, Ni-65, Np-237, Np-239, Pa-233, Pu-238, 

Pu-239, Pu-240, Pu-241, Pu-244, Ra-226, Ru-106, Th-229, Th-230, 

Th-232, U-232, U-232, U-233, U-235, U-236, U-238, Tc-99, Pa-231, 

Pu-242, Cm-244, Cs-137, Ni-63, Pu-241, Sr-90, Bk-249, Cf-250, U-234, 

Es-254

(b)

70A, Room 2229A Am-241, C-14, Fe-59, Np-237, Pb-210, Pu-238, Th-232, U-233, U-238, 

U-234/235

(b)

70A, Room 2229B Am-241, C-14, Fe-59, Np-237, Pb-210, Pu-238, Th-232, U-233, U-238, 

U-234/235

(b) 

70A, Room 4419 Kr-81, Kr-85 (b)

70A, Room 4429 Pb-205, Th-229, Th-232, U-233, U-238 (b)

70A, Room 4429A Pb-205, Th-229, Th-232, U-233, U-238 (b)

70A, Room 4429C Pb-205, Th-229, Th-232, U-233, U-238 (b)

70A, Room 4459 Th-228, Th-230, U(natural), U-234, U-235, U-238, Tc-99 (b)

70A, Room 4463 Th-228, Th-230, U(natural), U-234, U-235, U-238, Tc-99 (b)

71, Cave B (b) (b)

71, Cave N (b) (b)

72, Room 102 U-238 (b)

72, Room 112A U-238 (b)

72, Room 128 P-32, Na-21, Na-22, P-32, Co-57, Cu-60, Na-22, Ni-57, P-32, Sc-93, 

Zn-62, Zn-63, Ac-227, Am-241, Au-198, C-11, Cf-249, Cf-252, Co-56, 

Co-57, Co-58, Co-60, Cr-51, Cs-134, Cs-137, Eu-152, Eu-154, Fe-55, 

Fe-59, Mn-54, Na-22, Np-237, Np-239, P-32, Pa-233, Pu-238, Pu-239, 

Sb-124, Se-75, Zn-65, Be-7, Sc-46, Pa-231

(b)

72, Room 137 P-32, Na-21, Na-22, P-32, Co-57, Cu-60, Na-22, Ni-57, P-32, Sc-93, 

Zn-62, Zn-63, Ac-227, Am-241, Au-198, C-11, Cf-249, Cf-252, Co-56, 

Co-57, Co-58, Co-60, Cr-51, Cs-134, Cs-137, Eu-152, Eu-154, Fe-55, 

Fe-59, Mn-54, Na-22, Np-237, Np-239, P-32, Pa-233, Pu-238, Pu-239, 

Sb-124, Se-75, Zn-65, Be-7, Sc-46, Pa-231

(b) 

72C, Rooms 155, 

163, 169, 171, and 

173

U-238 (b)

73, Room 109 H-3 (b)

74 Co-60/P-32/Cs-137 1,500 (1962 and 1968)c

74, Room 144A Th-232, U(natural), U-238, Tc-99 (b)

74, Room 2011 Am-241, C-14, Fe-59, Np-237, Pb-210, Pu-238, Th-232, U-233, U-238, 

U-234/235

(b)

74, Room 238C H-3, O-32, P35 (b)

74, Rooms 265 

and 265A

P-32 (b)

74, Room 285 C-11, Ce-141, Cu-64, F-18, I-123, I-125, I-131, N-13, Nb-95, O-15, 

Ru-103, Sr-85, Tc-99m, Tl-201, Sc-46

(b)

74, Room 312 C-11, C-14, H-3, P-32 (b)

74, Room 330 C-11, C-14, H-3, P-32 (b)

74, Room 330A C-11, C-14, H-3, P-32 (b)

74, Room 344 C-11, C-14, H-3, P-32 (b)

74, Room 350 C-11, C-14, H-3, P-32 (b)

74D, Room 1 C-11, Ce-141, Cu-64, F-18, I-123, I-125, I-131, N-13, Nb-95, O-15, 

Ru-103, Sr-85, Tc-99m, Tl-201, Sc-46

(b) 

75 Cs-137 30 (March 1967)c

75, Room 103, 107 

and Storage 

Locker

H-3 (b) 

75, Room 109 H-3, activation products (b)

75, Room 113 H-3, activation products (b)

75, Room 127 Alpha, Am-241, beta-gamma, C-11, C-14, F-18, H-3, I-125, I-129, 

I-131, P-32, S-35, U-232, Sr-90

(b)
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 17 of 66

Building Radionuclides Activity (Ci)a

75D, Room 101B Activation Products, Various, Kr-81, Kr-85 (b)

76, Room 135 and 

Storage Locker

Am-241, C-11, C-14, F-18, H-3, I-125, I-129, I-131, P-32, S-35, U-232, 

Sr-90

(b)

84, Room 101 P-32 (b)

84, Room 153 C-14, H-3, P-32 (b)

84, Room 155 H-3, P-32, S-35 (b)

84, Room 157 H-3, P-32, S-35 (b)

84, Room 161 H-3, P-32, S-35 (b)

84, Room 175 H-3, P-32, S-35 (b)

84, Room 201 P-32 (b)

84, Room 220 P-32 (b)

84, Room 263 P-32, C-14, H-3 (b)

85, Room MW1 Alpha, beta-gamma, C-14, H-3 (b)

85, Room MW2 Alpha, beta-gamma, C-14, H-3 (b)

85, Room MW3 Alpha, beta-gamma, C-14, H-3 (b)

85, Room MW4 Alpha, beta-gamma, C-14, H-3 (b)

85, Room MW5 Alpha, beta-gamma, C-14, H-3 (b)

85, Room MW6 Alpha, beta-gamma, C-14, H-3 (b)

85, Room MW7 Alpha, beta-gamma, C-14, H-3 (b)

85, Room MW8 Alpha, beta-gamma, C-14, H-3 (b)

85, Room RW1 Alpha, beta-gamma, C-14, H-3 (b)

85, Room RW2 Alpha, beta-gamma, C-14, H-3 (b)

85, Room 

Scintillation Vial 

Room

Alpha, beta-gamma, C-14, H-3 (b)

85, Room Staging 

Area

Alpha, beta-gamma, C-14, H-3 (b)

88, Room 134 Ac-227, activation products, Am-241, Cf-249, Cf-252, Gd-148, Pb-212, 

Th-228, Th-229, Th-232, U(natural), MFPs, various, actinide tracers, 

gamma tracers

(b)

88, Room 135 H-3, C-11, N-13, Na-22, Be-7, O-14, Ne-19, Ne-18, Br-76, Br-77, Kr-76, 

Kr-77, Kr-79, P-32, Na-21, Na-22, P-32, Co-57, Cu-60, Na-22, Ni-57, 

P-32, Sc-93, Zn-62, Zn-63, Ac-227, Am-241, Au-198, C-11, Cf-249, 

Cf-252, Co-56, Co-57, Co-58, Co-60, Cr-51, Cs-134, Cs-137, Eu-152, 

Eu-154, Fe-55, Fe-59, Mn-54, Na-22, Np-237, Np-239, P-32, Pa-233, 

Pu-238, Pu-239, Sb-124, Se-75, Zn-65, Be-7, Sc-46, Pa-231

(b) 

88, Cave 0 Be-7, C-11, C-14, F-17, F-18, N-13, Na-22, O-14, O-15, Br-76, Br-77, 

Kr-76, Kr-77, Kr-79, Ne-19, Ne-18 

(b)

88, Cave 1 C-11, N-13, Na-22, Be-7, O-14, Ne-19, Ne-18, Ac-227, activation 

products, Am-241, Cf-249, Cf-252, Gd-148, Pb-212, Th-228, Th-229, 

Th-232, U(natural), MFPs, various, actinide tracers, gamma tracers

(b)

88, Cave 2 Ac-227, activation products, Am-241, Cf-249, Cf-252, Gd-148, Pb-212, 

Th-228, Th-229, Th-232, U(natural), MFPs, various, actinide tracers, 

gamma tracers

(b)

88, Cave 3 H-3, C-11, C-14, P-32, P-33 (b)

88, Cave 4 H-3, C-11, C-14, P-32, P-33 (b)

88, Cave 4C Activation products, Be-7, C-11, C-14, F-17, F-18, N-13, Na-22, O-14, 

O-15

(b)

88, Cave 4C roof Ac-227, activation products, Am-241, Cf-249, Cf-252, Gd-148, Pb-212, 

Th-228, Th-229, Th-232, U(natural), MFPs, various, actinide tracers, 

gamma tracers

(b)

88, Cave Roofs Act Products, Be-7, C-11, C-14, F-17, F-18, N-13, Na-22, O-14, O-15, 

Ne-19, Ne-18

(b)

88, East Alley Br-76, Br-77, Kr-76, Kr-77, Kr-79 (b)

88, East alley 

mezzanine

Br-76, Br-77, Kr-76, Kr-77, Kr-79, Ac-227, activation products, Am-241, 

Cf-249, Cf-252, Gd-148, Pb-212, Th-228, Th-229, Th-232, U(natural), 

MFPs, various, actinide tracers, gamma tracers

(b)
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 18 of 66

Building Radionuclides Activity (Ci)a

88, East alley niche C-11, N-13, Na-22, Be-7, O-14, Ne-19, Ne-18, Ac-227, activation 

products, Am-241, Cf-249, Cf-252, Gd-148, Pb-212, Th-228, Th-229, 

Th-232, U(natural), MFPs, various, actinide tracers, gamma tracers

(b)

88, Entire Vault 

roof

Activation products, Be-7, C-11, C-14, F-17, F-18, N-13, Na-22, O-14, 

O-15, Br-76, Br-77, Kr-76, Kr-77, Kr-79

(b) 

977, Room 205 H-3, P-32, P-33, S-35 (b)

977, Room 240 H-3, P-32, S-35 (b)

977, Room 285 C-14, Cd-109, H-3, P-32, S-35 (b)

a. Maximum quantity present in the building in curies unless otherwise noted. 

b. No quantity and/or isotope information available.

c. Date used where indicated in the historical documents.

3.0

Table 3-1 lists the entrance skin exposure (ESE) and organ doses for posterior-anterior (PA) chest 

radiography at different times at LBNL. Beginning in the early years and continuing through March 

1975, the assigned doses are the default values given by ORAUT (2005a). Calculated doses for the 

later times are based on measurements of the ESE and other factors with the LBNL equipment. The 

earliest document found (de Castro and Thomas 1975) reports results from a survey of the Picker 

radiographic X-ray unit conducted on March 19, 1975. The report states, “At a typical chest technique 

of 86 kV, 200 mA, large spot, 1/30 sec, 72 inches [= 183 cm], the direct beam exposure was 7 mR.” 

Collimation is described as “satisfactory,” and the filtration is “assumed sufficient [6].” For the PA view, 

if one assumes a chest thickness of 26 cm and allows 5 cm for the thickness of the cassette, the 

implied source-to-skin distance is 183 cm – 31 cm = 152 cm [7]. It follows from the measurement that 

the ESE is (7 mR)(183/152)2 = 10 mR. These data are consistent with subsequent survey results 

reported through 1986 for this X-ray machine (de Castro 1977; Vaughan 1982; de Castro 1983; 

de Castro and Hill 1984; de Castro 1986). This unit was in use until its replacement by a new 

machine about September 28, 1987 (Bradfield 1987). The change was made to a Picker model BGX 

625R stationary, general-purpose system, manufactured in 1987 (Thomas 1991; Bradfield-Montoya 

1994). Based on available documents, the two Picker machines apparently served in succession for 

making medical X-rays at LBNL dating back to at least 1975. Chest X-rays were discontinued at 

LBNL, and the second Picker machine was removed by about the end of 1993 [8]. No policy 

statements with records of frequency were found. In the absence of other information, a preemployment PA chest radiograph and subsequent annual frequency should be assumed [9]. 

OCCUPATIONAL MEDICAL DOSE

Under an interagency agreement, the second Picker system was given a thorough inspection by the 

Food and Drug Administration on February 23, 1989 (Goldstein 1989). All components were found to 

operate within applicable Federal or manufacturer’s specifications. The measured exposure for a 

routine chest X-ray was 11.2 mR with the technique of 300 mA at 84 kVp in the automatic exposure 

control mode. No significant problems were incurred in the operations. Regular control procedures 

were conducted to monitor performance over the years. Over time, the ESE measured in subsequent 

surveys remained close to 10 mR. Allowing a conservative factor of approximately 2, the value ESE = 

20 mR is used to estimate organ doses for all times after March 1975. 

In Table 3-2, the data from Table 3-1 are presented as the maximum organ doses in any calendar 

year. Following ORAUT (2005a), an uncertainty of ±30% at 1 sigma is estimated for an individual 

ESE or organ dose.
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 19 of 66

Table 3-1. Parameters and organ doses for 14- by 17-in. PA chest radiography.

a,b,c

Period Frequency Applicability

PA chest 

ESE (mR)

Organ doses (rem) from 14- by 17-in. PA chest radiography

Thyroid Ovaries Testes Lungs Breast Uterus

Bone 

marrow

Pre-1970 Unknown Unknown 200 3.48E-02 2.50E-02 5.00E-03 9.02E-02 9.80E-03 2.50E-02 1.84E-02

1970–3/75 Unknown Unknown 100 3.20E-03 1.00E-04 1.00E-06 4.51E-02 4.90E-03 1.30E-04 9.20E-03

4/75–1979 Unknown Unknown 20 6.40E-04 2.00E-05 2.00E-07 9.02E-03 9.80E-04 2.60E-05 1.84E-03

1980–1993 Unknown Unknown 20 1.56E-03 1.04E-04 2.00E-07 1.35E-02 2.32E-03 1.04E-04 3.56E-03

Period

Organ doses (rem) from 14- by 17-in. PA chest radiography

Eye/brain

Liver/gall 

bladder/spleen

Urinary 

bladder

Colon/ 

rectum Thymus Esophagus Stomach

Bone 

surface Remainder Skin

Pre-1970 6.40E-03 9.02E-02 2.50E-02 2.50E-02 9.02E-02 9.02E-02 9.02E-02 9.02E-02 9.02E-02 2.70E-01

1970–3/75 3.20E-03 4.51E-02 1.00E-04 1.00E-04 4.51E-02 4.51E-02 4.51E-02 4.51E-02 4.51E-02 1.35E-01

4/75–1979 6.40E-04 9.02E-03 2.00E-05 2.00E-05 9.02E-03 9.02E-03 9.02E-03 9.02E-03 9.02E-03 2.70E-02

1980–1993 1.56E-03 1.35E-02 1.04E-04 1.04E-04 1.35E-02 1.35E-02 1.35E-02 1.35E-02 1.35E-02 2.80E-02

a. Values for the pre-1970 and the 1970 to 3/1975 periods are taken from Table 6-5 of ORAUT (2005a).

b. The default values (ORAUT 2005a) for beam quality of 2.5 mm Al HVL prior to 1980 and 4.0 mm Al HVL for subsequent years are assumed.

c. No policy statement on LBNL frequency was found. In the absence of other information, the assumption of an annual frequency is reasonable.

Table 3-2. Maximum organ doses by year for 14- by 17-in. PA chest radiography. 

Year

Organ doses (rem) from 14- by 17-in. PA chest radiography

Maximum 

thyroid

Maximum 

ovaries

Maximum 

testes

Maximum 

lungs

Maximum 

breast

Maximum

uterus

Maximum bone 

marrow

Maximum 

eye/brain

Pre-1970 3.48E-02 2.50E-02 5.00E-03 9.02E-02 9.80E-03 2.50E-02 1.84E-02 6.40E-03

1971–1975 3.20E-03 1.00E-04 1.00E-06 4.51E-02 4.90E-03 1.30E-04 9.20E-03 3.20E-03

1976–1979 6.40E-04 2.00E-05 2.00E-07 9.02E-03 9.80E-04 2.60E-05 1.84E-03 6.40E-04

1980–1993 1.56E-03 1.04E-04 2.00E-07 1.35E-02 2.32E-03 1.04E-04 3.56E-03 1.56E-03

Year

Organ doses (rem) from 14- by 17-in. PA chest radiography

Maximum 

liver/gall 

bladder/spleen

Maximum

urinary 

bladder

Maximum 

colon/rectum

Maximum 

thymus

Maximum 

esophagus

Maximum 

stomach

Maximum bone 

surface

Maximum 

remainder

Maximum 

skin

Pre-1970 9.02E-02 2.50E-02 2.50E-02 9.02E-02 9.02E-02 9.02E-02 9.02E-02 9.02E-02 2.70E-01

1970–1975 4.51E-02 1.00E-04 1.00E-04 4.51E-02 4.51E-02 4.51E-02 4.51E-02 4.51E-02 1.35E-01

1976–1979 9.02E-03 2.00E-05 2.00E-05 9.02E-03 9.02E-03 9.02E-03 9.02E-03 9.02E-03 2.70E-02

1980–1993 1.35E-02 1.04E-04 1.04E-04 1.35E-02 1.35E-02 1.35E-02 1.35E-02 1.35E-02 2.80E-02
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 20 of 66

4.0

Occupational environmental dose results from releases or direct radiation due to facility operations. 

Inhalation of environmental radionuclides results in internal dose to the whole body or body organs. 

The site profile analysis determined the internal dose for workers outside the facilities from air 

concentrations from releases from stacks and individual buildings. 

OCCUPATIONAL ENVIRONMENTAL DOSE

The site profile preparers reviewed site environmental reports for data that would be useful in 

reconstructing ambient radiation levels. Ambient radiation dose rates include natural background 

radiation and sources at the facility. 

The laboratory used and created small amounts of radioactive material, created localized radiation 

areas with machines (X-ray machines and accelerators), and produced tritiated materials for wide 

distribution to other research facilities. Nuclear reactions were studied in the accelerators, which 

started on the site in the early 1930s as small devices (bench top) and progressed to specially built 

large buildings with controls and shielding. Workers and technicians who had access to the 

accelerators and experimental rooms had the highest risk and were, therefore, monitored for radiation 

exposure.

LBNL was involved in much of the discovery of the radiological properties of materials, and 

awareness of the hazards grew with each step in size and power of accelerator. Those working with 

the machines were monitored by state-of-the-art methods for internal and external dose [10]. The 

LBNL hillside site is small. The distance from the accelerators to the maximum exposed member of 

the public was less than 1,500 m. Documentation of the site boundary impact began in 1959 and 

continues today.

Tritium was used in large quantities in the National Tritium Labeling Facility (NTLF), which began 

operation in the early 1980s. Releases and doses to the personnel and public were well documented 

over the course of operations, and those practices continue today after shutdown. Potable water has 

always been supplied from off the site by the local utility.

Methods of measuring deposition and air concentrations have varied over the years in terms of 

technique and locations. A cursory review shows long-term interest in fallout. Fallout and naturally 

occurring radioactive materials dominate any deposition from onsite releases and resuspension. 

Radionuclides were either created on the site or purchased, and the large variety created an 

extensively diverse source term. However, except for tritium and accelerator-induced byproducts, 

these nuclides were present only in small quantities needed for research. Laboratories at LBNL were 

some of the first to use hoods and airflow to control exposure to hazardous material [11] (Thaxter 

1950a).

Interviews with the dosimetry staff indicated that every monitored worker has a dosimetry file that is 

accessible today. Those who were not monitored would have been exposed to external radiation 

from outside the well-designed shield walls and from releases through ventilation systems. In 

addition, these facilities were always being improved and were operated on an intermittent schedule. 

Concepts of long-term steady-state exposure are not consistent with the nature of this facility.

Table 4-1 lists the maximum external gamma and neutron radiation measurements in millirem per 

year. There is not sufficient data available to provide measured uncertainty. The values listed in the 

table are those documented in the annual LBNL environmental reports. Early interest was the 

maximum possible dose from surveys (usually during accelerator runs) and analysis. In later years 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 21 of 66

with better monitoring, improved shielding, and the development of as-low-as-reasonably-achievable 

(ALARA) policy, the reported doses fell dramatically [12]. The annual average measurements, as 

opposed to the maximum values, were at 50% of maximum or less [13]. It is reasonable to use 75% 

of the maximum as the likely value, and 25% of the maximum as the likely uncertainty as an estimate 

of unmonitored dose [14]. 

Table 4-1. External gamma and neutron radiation.

a,b,c

Year

Dose 

(mrem/yr) Reference

1942 140 --

1943 150 --

1944 170 --

1945 190 --

1946 210 --

1947 230 --

1948 250 --

1949 280 --

1950 310 --

1951 350 --

1952 390 --

1953 430 --

1954 480 --

1955 530 --

1956 590 --

1957 660 --

1958 730 --

1959 810 Patterson (1962)

1960 650 Patterson (1962)

1961 450 Patterson (1962)

1962 312 LRL (1963)

1963 173 Lichliter (1964)

1964 113 Patterson (1965)

1965 74.6 Patterson (1966)

1966 124 LRL (1967)

1967 76 LRL (1968)

1968 132 LRL (1969)

1969 291 Kelly (1970)

1970 173 LRL (1971)

1971 283 Kelly (1972)

1972 57 Cantelow (1973)

1973 28 Thomas (1974)

1974 28.3 Stephens and Cantelow (1975)

Year

Dose 

(mrem/yr) Reference

1975 95.87 Cantelow (1976)

1976 5.35 Stephens (1977)

1977 5.3 Stephens (1978)

1978 7.29 Schleimer (1979)

1979 10.1 Schleimer (1980)

1980 4 Schleimer (1981)

1981 15.1 Schleimer (1982)

1982 24.5 Schleimer (1983)

1983 5.8 Schleimer (1984)

1984 5.4 Schleimer (1985)

1985 1.8 Schleimer (1986)

1986 3.5 Schleimer (1987)

1987 3.5 Schleimer (1988)

1988 1.9 Schleimer (1989)

1989 2.6 Schleimer and Pauer (1990)

1990 7 Schleimer and Pauer (1991)

1991 2 Pauer, Schleimer, and Javendal 

(1992)

1992 2.3 Balgobin et al. (1993)

1993 2.96 University of California (1994)

1994 1.53 LBNL (1995a)

1995 2.1 LBNL (1996)

1996 3.3 LBNL (1997)

1997 2 LBNL (1998)

1998 5 LBNL (1999)

1999 4 LBNL (2000)

2000 4 LBNL (2001)

2001 0.65 LBNL (2002)

2002 3.2 LBNL (2003)

2003 0.2 LBNL (2004)

2004 0.28 LBNL (2005a)

a. No values were reported from 1942 to 1958; the values provided are projections.

b. Highest of perimeter or onsite data was used. In general, due to the terrain of the facility, the perimeter readings are higher than the 

local onsite readings. This is due to the small size of the site, elevation differences, and skyshine with more shielding in the walls than 

roofs of the accelerator buildings. The dose values reflect a 365-d 24-hr measurement less background [15]. 

c. Energies are 30 to 250 keV for photons and 0.1 to 2 MeV for neutrons. Neutron component of dose is typically ≥70% of total[16].

The following is a description of how the estimates of external exposure were generated for the period 

before 1959 and for two recent years (2002 and 2004) for which measurements of "not detected" 

were reported.

The 2002 and 2004 estimates were based on averages of the 5 preceding years rounded to two 

significant digits.

As the number and power of accelerators grew on the site, exposures to unmonitored workers and the 

environment grew (Goldhaber 1981). Many of the earlier reports would not meet the expectations of 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 22 of 66

today and often included values based on crude measurements and calculations with only enough 

precision to demonstrate that limits were not exceeded. The use of the first reported doses beginning 

in 1959 probably produces projected doses much greater than actual doses. These reported values 

form a basis for the dose estimates for the preceding years as described in the next three paragraphs

[17]. 

An assumption of linear growth from 1942 to 1959 would indicate under-reported doses in the late 

1950s, which would be expected to be closer to the doses in the 1960s [18]. An assumption of the 

average dose reported from 1959 to 2004 would overestimate the doses in the 1940s [19]. This 

analysis estimated the doses based on a compromise using a parabolic curve that begins near 

background in 1942 and peaks with the reported value in 1959 [20]. 

The parabolic curve estimates provide sufficient dose for the workers in the 1950s and more 

reasonable doses in the first years. The average value in these estimates (0.348 rem/yr) is four times 

the average reported (0.087 rem/yr) from 1959 to 2003. This provides an ample margin to ensure that 

the dose is not underestimated [21]. 

In Thomas et. al. (2000) it was determined that 1959 had the highest annual dose of any year at 

LBNL. In this report doses were revised based on modern knowledge of both the energy spectrum of 

accelerator produced neutrons and the appropriate coefficient functions for different irradiation 

geometries. For the years 1959 through 1975 the revised values were approximately one-half the 

values shown in Table 4-1. Values were also estimated back through calculation to 1954 when the 

Bevatron became operational. These estimated values are consistently lower than those shown in 

Table 4-1. Using the higher values shown in Table 4-1 is favorable to claimants without being 

excessively conservative, i.e. a factor of about two times.

Table 4-2 lists the maximum site-wide annual median intakes in becquerels per year via inhalation. 

The data were derived from the LBNL annual environmental reports; the maximum reported values 

were used. Water supplies are from East Bay Municipal Utility District off the site, and no indication of 

soil contamination exceeding a background level has been found, so there would have been no doses 

from intakes from soil and drinking water. LBNL attributes gross alpha results to 232Th and gross beta 

results to 90Sr for the purposes of dose calculation; these assumptions should be made. Thorium is 

representative of naturally occurring alpha emitters, and strontium is representative of the beta 

component of fallout. By using these nuclides in the calculations, the resulting doses would not 

underestimate the actual doses [22]. 

Missing data are filled in with the estimates as explained below and are based on the available 

maximum observed values in the annual environmental reports.

• Alpha: Reported alpha concentrations appear to be independent of operational effects of the 

Laboratory and relatively constant over time. Therefore, a good representation of the alpha 

concentrations and intake is the average of all the measurements, which results in an intake of 

0.43 Bq/yr [23]. 

• Beta: The principal contributor to gross beta concentrations is fallout from atmospheric 

nuclear weapons testing. The later years of missing data were assigned a value of 10 Bq/yr 

for gross beta intake [24]. This represents a modest overestimate of the intake. The intakes 

in the early years before 1961 were estimated to be reduced at a rate that would bring the 

intake in 1942 to values close to the most recent observations. This would account for the 

gradual rise in beta concentrations due to weapons testing and the effects (if any) of increased 

nuclide inventory on the site and increased power levels of the accelerators [25]. 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 23 of 66

• Tritium: Tritium releases, concentrations, and subsequent intakes are related to accelerator 

and NTLF operations. Annual data are available from 1974 to 2003. The value for 2004 

Table 4-2. Maximum site-wide annual median 

intakes (Bq/yr) via inhalation.

a,b,c,d

Year

Respiration rate: 2,400 m3/yr (2,000 hrs/yr)

Gross

alpha

Gross

beta Tritium C-14

1942 0.43 12 4,900 760

1943 0.43 15 5,800 890

1944 0.43 17 6,800 1100

1945 0.43 20 8,000 1200

1946 0.43 24 9,400 1500

1947 0.43 28 11,000 1700

1948 0.43 33 13,000 2000

1949 0.43 39 15,000 2400

1950 0.43 46 18,000 2800

1951 0.43 54 21,000 3300

1952 0.43 63 25,000 3900

1953 0.43 75 29,000 4,500

1954 0.43 88 34,000 5,300

1955 0.43 100 40,000 6,300

1956 0.43 120 48,000 7,400

1957 0.43 140 56,000 8,700

1958 0.43 170 66,000 10,000

1959 0.43 200 78,000 12,000

1960 0.43 230 91,000 14,000

1961 0.3432 273.6 110,000 17,000

1962 0.0888 511.2 130,000 20,000

1963 0.0888 849.6 150,000 23,000

1964 0.0888 198.96 170,000 27,000

1965 0.0888 35.52 210,000 32,000

1966 0.0888 12.432 240,000 38,000

1967 0.0888 29.28 280,000 44,000

1968 0.0888 28.32 330,000 52,000

1969 0.0888 17.76 390,000 61,000

1970 0.444 44.4 460,000 72,000

1971 0.444 23.088 550,000 85,000

1972 0.0888 7.104 640,000 100,000

1973 0.07104 1.8648 750,000 120,000

1974 0.444 32.856 888,000 17,760

1975 0.6216 26.4 213,120 30,240

1976 0.5328 71.04 328,800 239,760

1977 0.3552 46.08 710,400 79,920

1978 0.444 24.96 2,222,400 40,080

1979 0.6216 20.424 1,687,200 32,856

1980 0.5328 21.312 62,160 31,080

1981 0.444 41.76 97,680 17,760

1982 0.3552 12.432 266,400 26,640

1983 0.1776 9.768 355,200 26,640

1984 0.2664 12.432 8,880,000 2,664,000

1985 0.3552 12.432 3,720,000 97,680

1986 0.07104 4.44 1,065,600 35,520

1987 0.444 17.76 444,000 35,520

1988 0.444 14.208 266,400 17,760

1989 0.7992 11.544 976,800 26,640

1990 0.5328 12.432 266,400 35,520

1991 0.444 1.5984 444,000 17,760

1992 0.5328 15.096 631,200 17,760

1993 0.444 14.208 427,200 31,200
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 24 of 66

Year

Respiration rate: 2,400 m3/yr (2,000 hrs/yr)

Gross

alpha

Gross

beta Tritium C-14

1994 3.6408 6.216 213,120 142,080

1995 0.43 10 184,800 8,160

1996 0.43 10 13,200 2100

1997 0.43 10 26,640 4100

1998 0.43 10 21,384 3300

1999 0.552 3.12 2,592 2,592

2000 0.43 10 12,168 1900

2001 0.43 10 11,208 1700

2002 0.43 10 2,952 500

2003 0.43 10 528 100

2004 0.48 3.6 0 384

a. The site is so small that the environmental samples are 

representative of the site as a whole.

b. Highest of perimeter or onsite data was used. In general, due to 

the terrain of the facility, the perimeter readings are higher than the 

local onsite readings due to the small size of the site, elevation 

differences, and the height of the stack release points. Since 

1964, the maximum reported value for the year was used except 

for 1986 because of the inordinate contribution from Chernobyl. 

Earlier years showed significant fluctuation due to fallout; because 

of its location on the west coast, LBNL routinely transmitted data 

on fallout to the U.S. Atomic Energy Commission.

c. Estimated values in italics

d. Data from environmental reports listed in Table 4-2 for the 

appropriate year. The references are included in the References 

section of this document and are available from the Site Research 

Database or at www.lbl.gov.

(2,500 Bq/yr) was chosen to conform with the trend in previous years and average ratio to 14C 

(see 14C discussion below) [26]. The years before 1974 were trended downward back to 1942 

to values consistent with recent measurements, which are the lowest values observed. The 

assumption is that the site tritium concentration before operations began was no higher than 

the lowest observed concentrations. This is consistent with the trend of accelerator 

development and the operation of the NTLF [27]. 

• Carbon-14: Carbon-14 releases, concentrations, and intakes are related to accelerator 

operations. The average ratio of 14C to tritium in the years during which both were reported is 

0.155 [28]. When one data point is missing, this value is used to estimate missing data in later 

years. For the years before 1974, a trend identical to the tritium estimates was employed.

All estimates were limited to no more than two significant digits.

5.0

The LBNL bioassay records show that the selection of personnel for bioassay and the radionuclides 

for analysis have been based on the work performed by the individual. Selection of employees to be 

included in the bioassay program was typically made by the Laboratory’s Safety Services Department 

through its staff of monitors. The monitors were directly aware of the radionuclides used throughout 

the Laboratory and were therefore best qualified to select employees at risk for potential internal 

exposure. Unmonitored workers should be assessed using coworker data and the tolerance levels 

discussed below if there is evidence that they worked with uncontained radioactive materials; 

otherwise, environmental data should be used.

OCCUPATIONAL INTERNAL DOSE
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 25 of 66

Bioassay measurement results were noted in the files as early as 1946. The earliest bioassay results 

are in vitro measurements. In vitro samples were primarily routine urine samples but even the earliest 

records showed measurement results for feces and sputum. Methods used to analyze the samples 

varied but relied primarily on standard methods (McClelland 1958) in use at other laboratories such as 

Los Alamos National Laboratory (LANL). Bioassay monitoring programs were in place beginning in 

1947. However, in the early years monitoring was performed on an as-needed basis and was not 

covered by a formal documented program (LRL undated). Routine bioassay monitoring of employees 

was initiated about 1961 (Howe 1961a). Beginning in 1996, the bioassay program changed such that 

personnel were selected for operational bioassay based on the radionuclide authorization program 

and reviews of work performed. The program currently uses these operational bioassays to evaluate 

worker exposure or, most often, to verify the absence of internal exposures. This program was 

started because so few worker groups were expected to exceed the 100 mrem/yr of internal exposure

required to put them into a routine monitoring program [29]. 

Table 5-1 lists the in vitro types of bioassay, the periods, and the frequencies of monitoring. Table 5-2 

lists the in vitro sample types and the analysis codes found in the records. 

From 1960 to 1996, both in vitro and in vivo monitoring records and associated interpretations exist 

and are available in the spreadsheet discussed below. In vivo monitoring consisted primarily of 

whole-body counts (WBCs) performed at LBNL from 1960 to 1996. The WB counter used throughout 

this period was at the Donner laboratory. It was designed, built, and operated by Dr. Thornton 

Sargent, III, who modeled it after the counter at Argonne National Laboratory–East (ANL-E). The 

WBC was in an 8- by 9- by 6-ft room, shielded by walls of 6-in. “pre-atomic age” steel lined with 

0.125 in. of lead. The main detector was a 9.375- by 4-in. NaI(Tl) crystal positionable in any geometry 

over a 1-m arc bed or an Argonne chair. The large crystal was used for counting energies above 

50 keV. The WBC was equipped with a probe that could be mounted with 2-in. diameter crystals 

either 0.125 or 0.25 in. thick for counting low-energy gamma-emitting isotopes such as 125I. The 1960 

calibration of the counter was performed by “performing studies in which patients were injected with 

known amounts of various radioactive isotopes, and then immediately counted...” (Vargha 1996a).

After 1996, LBNL no longer performed WBCs but rather relied on Lawrence Livermore National 

Laboratory (LLNL) for occasional monitoring. Table 5-3 lists the in vivo monitoring and data storage 

from 1960 to the present.

Table 5-4 summarizes the detection limits found in available documents for in vitro bioassays. The 

minimum detectable activities (MDAs) in Table 5-4 were taken from the analysis results. In some 

cases, the analysis method was not available and the detection limit was not specified. Some of the 

records reviewed showed results for any positive net measurement, some simply noted "neg." LBNL 

appears to have reported only measurements above the detection limit. Table 5-6 lists the method 

codes used to identify analysis methods for bioassay samples. These codes were obtained from a 

computer listing of the analysis results [30]. 

Table 5-5 summarizes the detection limits found for in vivo bioassays. Because the WB counter was 

the same system for most of the history for LBNL, there is not much variation. Count times were 

typically 15 min but individual counts might have been increased to obtain lower detection limits [31]. 

The count time is noted in the individual records.

Copies of results from all individual bioassays were distributed to the employee, the medical files 

(Medical Services) and Safety Services. Vargha (1996b) contains information on the records for the 

WBCs between 1960 and 1996 and notes that the electronic files from 1960 to 1996 were available. 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 26 of 66

Table 5-1. Internal dose control program (in vitro). 

Monitoring type Routine/special Period Frequency

Urine, feces, and sputum from 1 employee assayed for alpha countsa Special 1946 (b)

Urine, feces, and blood from a few employees were sent to ANL-E for 

analysisa

Unknown, probably 

special

1947 (b)

All employees included in medical program that included chest X-ray, 

hematology, and urinalysisa 

Routine 1947 Periodic

Began developing a bioassay program through LANLa Unknown 1950 (b)

Samples were sent to LLNL for analysisa Unknown 1956–1959 (b)

Bioassay laboratory was started and presumably routine urine samples 

were analyzeda

Routine 1960 (b)

Nasal smears appeared to be obtained as a matter of routine following 

spills or after hood filter changesc

Special 1957–1960 (b)

Urine 24-hr, employees working with transuranic elements, Sr-90, and 

radiumd

Routine 1961 Quarterly

Urine 24-hr, employees working with activity but not defined by the 

quarterly frequencyd

Routine 1961 Annual

Urine 24-hr, employees with less potential for exposured Routine 1961 Every 5 yr

Urine 24-hr, radiochemists and Health Chemistry personnel more highly 

exposed to alpha emitterse

Routine 1962 Semiannual

Urine 24-hr, radiochemists and Health Chemistry personnel, selected 

members of building trades frequently assigned work in active areas, and 

selected administrative personnel whose exposure is essentially zero and 

who should be considered controlse

Routine 1962 Annual

Urine 24-hr, Sr-90f Special 1962 Weekly

Semiweekly

Feces [32] Special 1962 Special

Blood [33] Special 1962 Special

Urine 24-hrg Routine 1974 Annualc

Urine, feces, sputumg Special 1974 Specialc

Urineh Routine 1975 Annual

Fecesh Special 1975 Special

Breath, radioactive carbonh 1975 Special

Urine spot, H-3 [34] 1983 (b)

Urine spot, C-14 1993 (b)

Urine, 24-hr [35] Routinei 1995–present (b)

Feces [36] Speciali 1995–present (b)

Urine, spot [37] 1995–present (b)

a. LRL (undated).

b. Frequency not applicable or not available.

c. Nasal smear alpha and beta counts were documented following numerous spills in Building 70 between 1957 and 1961. The 

radionuclides involved in the spill were sometimes documented: Pu, Am, Pu-240, Pu-239, Am-241, Cm-244, Ru-106, Ac-227, and Tb161 (Kaufman 1957; Alloway 1957; LRL 1958a,b,c,d, 1959a,b, 1960, 1961). 

d. Howe (1961a) identifies that LBNL should establish a “full-fledged bioassay program.” An attachment to the memorandum, “Bioassay 

Program Features,” identified routine scheduling. It is not clear if the quarterly frequency was implemented. Review of an electronic 

file of historical samples shows a number of employees provided routine urine samples more frequently than annually but it might have 

been at the discretion of the laboratory staff based on type of work rather than a scheduled frequency. 

e. Soule (1962).

f. Selected individuals performing Sr-90 work were followed at weekly or semiweekly intervals from March 1962 through the end of 1962 

(Low-Beer 1963). It is not clear that this was the practice throughout the history of the Laboratory. 

g. Hartsough (1974) indicates that the frequency for bioassay can be increased for employees who show a consistent history of positive 

bioassay results. Special monitoring was performed at the request of the employee or Safety Services of Health Physics. Examples of 

cause for special bioassay included fire involving radioactive materials, rupture of containment devices, air sample results at or above 

the maximum permissible concentration (MPC), high surface-swipe activity, and skin cuts or punctures while working with radioactive 

material. 

h. Pickler (1975).

i. Routine measurements are required only when the relative hazard of an operation shows a likelihood of incurring an intake that would 

result in a dose greater than 100 mrem. LBNL uses routine monitoring but relies mainly on “supplementary” bioassay programs. 

Routine programs are implemented when the dose equivalent is expected to be above 100 mrem. LBNL uses the quantity thresholds 

published in NUREG-1400 for relatively heavy elements (Hickey et al. 1993), and the quantity thresholds in the U.S. Nuclear 

Regulatory Commission Regulatory Guide Series for H-3 and radioiodines, to establish the need for routine bioassays. Supplementary 

bioassays are implemented at thresholds below the 100-mrem guideline. Threshold values used at LBNL are provided in AlMahamid 

(2005).
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 27 of 66

Table 5-2. Sample type (in vitro).a

Routine 

monitoring type Sample type Period Frequency

Urine 24-hr RU = Routine urine 1960–1997 Annual

Urine SU = Special urine 1960–1997 (b)

Blood SB =Special blood 1965–1988 (b)

Fecal SF = Special fecal 1960–1980 (b)

Nasal SN = Special nasal 1957–1960 

1969–1989

(b)

Sputum SS = Special sputum 1967–1974 (b)

Urine TU = Tritium urine 1968–1995 (b)

a. SB, SN, SS, and TU sample types should be verified. These codes are 

seen throughout the history of the records. In some cases the code is a 

best guess based on Szalinski (2006) and review of sample analysis. 

Two other codes (SH and SL) were noted in the files only once each and 

are therefore not in the table. The periods are generally taken from the 

historical database of records. 

b. Frequency not applicable or not available.

Table 5-3. Internal dose control program (in vivo). 

Monitoring type and storage Period Frequency

WBC–a

Records storage media:b

1. Alphabetically by individual, spectrum printouts are on 

paper tape

2. Electronic magnetic tape cassettes

Routine 1960–1983 Annual

Accident 1960–1983 (c)

Outside referral 1960–1983 (c)

Health chemistry 1960–1983 Annual

WBC–a

Records storage media:b

1. Alphabetically by individual

2. Electronic 8-in. diskettes

Routine 1983–1996 Annual

Accident 1983–1996 (c)

Outside referral 1983–1996 (c)

Special studies 1983–1996 (c)

WBC at LLNLd Special 1996–present (c)

a. WBCs were performed at LBNL from 1960 to 1996. 

b. Vargha (1996b).

c. Frequency not applicable or not available. 

d. AlMahamid (2005, Appendix 2). LBNL employee Jim Floyd indicated that after 1996 WBCs were no longer performed at 

LBNL and that LBNL relied on LLNL (Szalinski 2006). The Laboratory also moved away from routine annual monitoring 

around 1996 and relied on the authorization process for use of radionuclides and knowledge of the work to specify 

bioassay requirements and follow-up or periodic bioassays to verify the absence of internal exposures (Szalinski 2006). 

Since 1946 air sampling was performed routinely using the “filter queen approach” (Thaxter 1950b), 

gloveboxes to contain contamination were in routine use (Browne 1950; Thaxter 1950b, 1953, 1958), 

and air samples showed activity levels that were low compared to the permissible levels (Thaxter 

1958). This 1958 document noted that between 1948 and 1958 breathing zone air samplers were 

assessed daily in all operating laboratories and only an average of 3/mo of approximately 2,000 

monthly air samples showed air concentrations near the maximum permissible concentration (MPC)

for occupational exposure found in NCRP Handbook 52 (NCRP, 1953). Also noted was that between 

1948 and 1958 no air samples had significant airborne concentrations of long-lived alpha emitters 

approaching ten percent of the amount indicated as contributing a 1-wk dose to man (300 mrem/wk). 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 28 of 66

Table 5-4. In vitro detection limits.

a

Radionuclide Period

Sample 

typea Method/description MDA

Gross alpha 1957–1961 N Wet ashed (HNO3 + H2SO4) 60-min count time.b Method use of 3N HCl is also 

noted in 1957 records (Alloway 1957).

1 dpm

Gross alpha 

Isotopes of thorium, 

plutonium, curium, 

actinium, and

neptunium

1960–1969 U Livermore method without modification.c,d Coprecipitation of the activity with 

BiPO4.

Sometime later the method was changed to lanthanum fluoride coprecipitation 

and the MDA improved to 0.2 dpm/24-hr urine.

e

Method descriptions note that uranium, radium, and polonium are not detected 

by this method (Author unknown, no date). 

(Method Code 10)

0.3 dpm/24-hr 

urine 

(0.15 pCi/24-hr 

urine)

Gamma emitters

Gross alpha (assumed to 

include isotopes of 

thorium, plutonium, 

curium, actinium, and 

neptunium)

Gross beta (assumed to 

include Sr/Y-90, Sr-89, 

Ba/La-140, Ce/Pr-144, 

and fission products) 

1969–present U or F Analysis of 24-hr urine sample.f Preparation of the sample by alkaline phosphate

precipitation to eliminate monovalent cations. The ash from the alkaline 

phosphate precipitation is suspended in 10 mL of 2N HNO3. The whole sample 

is counted by gamma spectroscopy using a 4-in. NaI crystal and a 400-channel 

PHA (gamma: Method Code 04). One-fifth of the sample is then plated on 

aluminum for counting gross beta activity in a Nuclear-Chicago gas-flow 

proportional counter (gross beta: method code 50). The remaining four-fifths of 

the sample is processed by bismuth phosphate and lanthanum fluoride 

coprecipitation and counted for gross activity in the Nuclear-Chicago gas-flow 

proportional counter (gross alpha: Method Code 10).

Gamma 10 nCi

Alpha 0.1 pCi

Beta 1 pCi

Gross beta 1957–1961 N Wet ashed (HNO3 + H2SO4) 60-min count time.b Method use of 3N HCl is also 

noted in 1957 records (Alloway 1957).

30 dpm

Gross beta, Sr/Y-90, 

Sr-89, Ba/La-140, 

Ce/Pr-144, and fission 

products

1960–1969 U LANL method without modification. Phosphate precipitation. The method in 

McClelland (1958) describes that radionuclides include Sr/Y-90, Sr-89, 

Ba/La-140, Ce/Pr-144, and fission products. Another reference (Author 

unknown, no date) notes that protactinium is not detected by this method.

(Method Code 50)c,d

5.1 dpm/24 hr 

urine (1968)

2 dpm/24 hr 

urine (1969)

Am-241, Pu-239 1962 Column chromatography, electroprecipitation, and PHAg 1962 was the only 

reference to this method.

Not specified

Am-241, Am-243 1995–present U Alpha spectrometry 0.02 pCi

1995–present F Alpha spectrometry 0.02 pCi

Bk-249 1995–present U Gross beta or alpha spectrometry 0.02 pCi

C-14 1968–1993 U Sample is decolorized using activated charcoal and counted on a liquid 

scintillation counter. 

(Method Code 02)

5.0 nCi/L (1974)

1993–present U C-14 was listed in LBNL (1995b) but the method was not specified. 2.0 nCi/L (1993)

Cm-243, Cm-244, 

Cm-246, Cm-248, 

2005–present U or F Method not specified. 0.02 pCi/sample

Cf-252 1965–1995 U or F Gross alpha using LLNL method: Coprecipitation of the activity with BiPO4. 0.3 dpm/24-hr 

urine 

Cf-252, Cm-244 1968 Alpha PHA was used to indicate the relative ratio of Cf-252:Cm-244. Not specified
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 29 of 66

Radionuclide Period

Sample 

typea Method/description MDA

Cf-249, Cf-252 1995–present U or F Method not specified. 0.02 pCi

Fe-55, Fe-59 1995–present Method not specified. Not specified

H-3 1968–1982 U Liquid scintillation counting (Method Code 03) 0.02 μCi/L

1982–1995 U 0.01 μCi/L

1995–present U 4.5 pCi/sample

I-125 1980–1984 U Gamma spectroscopy (Method Code 05) 0.02 nCi/L

Na-22, Na-24 1995–present Method not specified. Not specified

Np-237, NP-239 1995–2005 U Method not specified. 0.05 pCi of 

Np-237

2005–present U or F Method not specified. 0.02 pCi/sample

P-32, P-33 1963–1995 U Ammonium phosphomolybdate precipitation (Method Code 71) 20 pCi/L

1995–present U or F Method not specified. 1.5 pCi

Protactinium 1960 Method not specified (1960 only year noted for protactinium). Not specified

2005–present 

(Pa-231)

U or F Method not specified. 0.02 pCi/sample

Pu-238, Pu-239 1995–present U or F Method not specified. 0.02 pCi

Pu-241 2005–present U or F Method not specified. 0.02 pCi/sample

Rare earths (primarily 

Pm-147)

1961 U Method not specified (1961 only year noted). 3 dpm/24-hr 

urine

Ra-226 1995–present U Method not specified. 0.1 pCi/L not 

listed in LBNL 

1995b but 

assumed h

Sulfur

S-35

1962–1992 U or F Urine analysis and fecal analysis

(Method Code 52)

0.2 nCi/L (1982)

1992–present U Barium sulfate precipitation 0.1 nCi/L (1992)

Strontium

Sr-90, Sr-89

1961–1995 Oak Ridge method of ion exchange chromatography (Sunderman and Townley 

1960)

(Method Code 51)

Not specified

Thorium 1995–2005 U Method not specified. Not listed in 

LBNL 1995b 

assume 0.1 

pCi/Lh

2005–present U or F Method not specified. 0.02 pCi/sample
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 30 of 66

Radionuclide Period

Sample 

typea Method/description MDA

Uranium 1961–1988 U LANL method of solvent extraction with dibutyl phosphate (McClelland 1958)

(Method Code 11)

0.15 dpm/24-hr 

urine

1988–1995 U Anion exchange chromatography 0.15 dpm/24 hr 

urine

1995–2005 U Method not specified. Not listed in 

LBNL (1995b); 

use previous 

MDA 0.15 

dpm/24-hr urine.

1995–2005 F Method not specified. Not listed; use 

1 pCi/sampleh

2005–present U or F Method not specified. 0.02 pCi/sample

a. Blanks in this table indicate not applicable or no data available. Sample types: N= nasal smears, U=urine, F=Fecal.

b. LRL (1958b).

c. Low-Beer (1962) and McClelland (1958).

d. Howe (1961b).

e. Author unknown (no date).

f. Patterson, Low-Beer, and Sargent (1969). Although this 1969 reference specifies a 10-nCi detection limit for gamma emitters, numerous bioassay records on the NIOSH-Office 

of Compensation Analysis and Support Claims Tracking System (NOCTS) showed a gamma detection limit of 30 dpm/specimen (13 pCi/specimen). Personnel performing dose 

reconstruction should use the MDA specified with the results.

g. Low-Beer (1963).

h. When an MDA was not specified for the radionuclides listed in the LBNL TBD for internal dosimetry (LBNL 1995b), the value from ANSI HPS N13.30, Table C.4 was used and 

assumed to be conservative. 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 31 of 66

Table 5-5. In vivo detection limits.

a

Radionuclide Method/description Period MDA

Gamma-emitting 

radionuclides 

50 keV–2 MeV

(including thyroid counts for 

I-131, I-123)

Be-7, Na-22,24, Mn-54, 

Ni-57,63, Fe-55,59, 

Co-57,58,60, Zn-65

WBC.b The Donner WB counter was an ANL-E type, with a 6-in. steel shield. 

The individual sits in an inclined chair and is counted by a 9- by-4- in. by 

9.375-in. NaI(Tl) crystal. The gamma pulses were analyzed by a 400-channel 

PHA, calibrated at 5 keV/PHA, for a 0–2.0 MeV spectrum range. Between 

1960 and 1983, data were recorded on printed paper tape, which was then 

keypunched for computer analysis. Calibration of the area under the 

photopeaks to yield microcuries of body burden was obtained from the spectra 

of medical patients injected with known quantities of short-lived radionuclides 

with various gamma-ray energies. Count time was routinely 15 min, but later 

references showed longer count times when lower MDAs or better counting 

statistics were needed and tolerated by the subject.

1960–1996 1–10 nCi for the 

nuclides listed

Assume 1 nCi for 

I-123 and I-125.

Assume 10 nCi 

for other listed 

nuclides.

Gamma-emitting 

radionuclides <50 keV (e.g., 

I-125)

WBC when count was performed using smaller (2-in.-diameter by 0.25-in.-

thick)b detector.

WBCs were also noted to identify the presence of high-energy beta emitters 

such as P-32 and Y-90, presumably through increases in the Compton 

continuum due to bremsstrahlung.c 

1960–1996 10 nCi for I-125d

Am-241, Am-243 Lung count for Np-239 [38] 1995–1996 50 pCi

F-18 WBC [39] 1995–1996 640 pCi

P-32 WBC (only one count for P-32 found in NOCTS) [40] 1971–1996 1E5 pCi

a. Blanks in this table indicate not applicable or no data available.

b. Patterson, Low-Beer, and Sargent (1969). 

c. Vargha (1996a).

d. A specific MDA for I-125 was not found. Vargha (1996b) specifies that the recommended count times produced MDAs for most isotopes in the range of 1 nCi. Based on the 

energy and photon yield for I-125, 10 nCi was assumed here.
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 32 of 66

Table 5-6. Unit codes and description of units [41]. 

Computer code

(Method Code) Description of units

Radionuclides reported 

with results

1 dpm/sample None listed

2 nCi/L 

dpm/g C (only one sample result observed was reported in dpm/g C)

C-14

3 μCi/L

nCi/L (only a few samples were reported in nCi/L

H-3

4 nCi/L I-125 (gamma emitters)

5 nCi/L I-125, Sr-85, Cr-51

pCi/L Cs-137

6 nCi/L I-125

10

dpm/d Gross alpha

dpm/L (one sample) Gross alpha

μCi/L (one sample) H-3

11 dpm/d U-238

14 Dpm None listed

15 dpm/d None listed

16 (a)

17 (a)

18 (a)

19 (a)

50 dpm/d, dpm/L Gross beta

51 dpm/d Sr-90

52 nCi/L S-35

54 (a)

55 nCi/L Ca-45

71 nCi/L P-32

a. Codes 16, 17, 18, 19, and 54 were observed in a computer listing of assays but there were no results or radionuclides reported with 

these codes. Note: only positive results are shown in the computer listing, copies of original records are marked “neg” when the result 

is below the detection level.

Additional table notes:

1. The available computer listing of results shows only measurement units for positive results and only radionuclides for some of the 

positive results.

2. These Method Codes are seen throughout the bioassay records from 1960 to 1996.

The Laboratory’s work was devoted to fundamental research on new elements, new isotopes, or 

properties of already known isotopes (Thaxter 1958). An important consideration at LBNL was that 

contamination could invalidate weeks of expensive research work and nuisance contamination was 

several orders of magnitude lower than the MPCs (Thaxter 1953). Air sample filter paper analysis 

results were compared to published tolerance levels for the radionuclide of concern (Thaxter 1950b). 

Tolerance levels were the 1949 Chalk River Tolerances (MPCs for radionuclides in air) (Saunders 

1950). A table of the 1949 MPCs from the Chalk River Canada conference (Warren 1949) is provided 

in Table 5-7.

Similar documentation of the low airborne concentration levels can be found in a 1953 document that 

refers to the previous 6 yr of air sample data (back to 1947) (Thaxter 1953). This document noted 

that daily breathing zone air samples outside the gloveboxes between 1947 and 1953 showed all 

samples to be below 5 × 10-11 μCi/cm3 beta/gamma. This document also noted the vigorous attempts 

to achieve no detectable airborne contamination (~2 dpm/ft3

) for long-lived transuranic alpha emitters.

Using the above information for long-lived radionuclides and noting that no air samples between 1948 

and 1958 had activity greater than 0.1, times the amount indicated as contributing a 1-wk dose 

(Thaxter 1958). Where the weekly dose was 0.3 rem, and, based on MPCs at the time, the intake of 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 33 of 66

Table 5-7. 1949 Chalk River tolerances, maximum 

permissible concentrations in inspired air.

a

Radionuclide

Maximum permissible 

concentrations in air

μg/cc μCi/ml

Ra-226 4.0E-12 4.0E-12

U-nat 25 μg/m3 b 1.7E-11

U-233 soluble salts 6.0E-9 6.0E-11

U-233 insoluble salts 2.5E-11 2.5E-13

Pu-239 5.0E-12 3.0E-13

Sr-90/Y-90 - 1.0E-10

H-3 - 1.0E-6

C-14 - 1.0E-6

Na-24 - 1.0E-6

P-32 - 2.0E-6

S-35 - 1.0E-6

Ar-41 - 1.0E-6

I-131 - 1.0E-9

Co-60 - 2.0E-9

Xe-133 - 1.0E-5

Xe-135 - 3.0E-5

a. Table values were obtained from Warren (1949), "Minutes of the 

Permissible Doses Conference held at Chalk River, Canada, 

September 29-30, 1949." In some cases the tolerance value was 

provided in μg/cm3 and was converted for this table to μCi/mL. 

b. U-nat was the one concentration reported in units of μg/m3.

227Ac and its progeny that will result in less than 0.3 rem/wk is 1.4 × 10-2 μCi. The maximum air 

concentration for long-lived radionuclides is calculated to be: 

uCiE ml

Ehrmwkhr ml m

wkuCiE /113

/42*/60*/40

/24.1*1.0 −= − 1

Using the information in Thaxter (1953), an upper limit on the beta/gamma airborne concentration 

could be assumed to be less than 5 × 10-11 μCi/cm3

. 

Air sample analysis detection limits for alpha particles were documented to be 1 cpm above 

background (Thaxter 1952). The counter efficiency is assumed to be 38% and the filter paper 

efficiency is assumed to be 75% (Thaxter 1950b).

Table 5-8 lists radionuclides used in each facility. Information on the compounds was very limited and 

therefore not included in the table. A few WBC records contained some information about the 

compound, but most were noted with “unknown” under the compound heading. The solubility type 

(type F, fast absorption, type M, moderate absorption, or type S, slow absorption) for some of the 

radionuclides can be found in International Commission on Radiological Protection (ICRP) Publication 

78, Individual Monitoring for Internal Exposure of Workers (ICRP 1998). Publication 78 also provides 

information on the class assigned for gases and vapors (Class SR-1 soluble or reactive, Class SR-2 

highly soluble or reactive, and Class SR-0 insoluble and nonreactive). The particle size for all 

 

1 The use of 227Ac and its progeny in this calculation as representing long lived isotopes comes directly from Thaxter 

(1958). Thaxter (1958) described a review of approximately 2000 air samples from the period 1948 - 1958 and that no air 

samples had approached 1/10 of the value representing 0.3 rem/week. It was described that 1.4 uCi of Ac-227 + daughters 

inhaled in a week results in a dose of 0.3 rem. This was then converted to an air concentration in this equation.
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 34 of 66

particulate radionuclides should be assumed to be 5-μm AMAD in accordance with the ICRP 

recommendations for occupational exposure.

Table 5-8. Radionuclidesa,b and fraction activitya by facility. Current 

building no. Building activity and compounds Radionuclide Fraction

1 Donner Laboratory (1961–present) C-14 4.28E-01

H-3 2.29E-01

I-125 2.31E-01

P-32 2.57E-02

S-35 8.55E-02

Alpha 1.41E-04

Beta 8.44E-04

3 Calvin Laboratory C-14 4.40E-03

H-3 5.37E-02

P-32 8.59E-02

P-33 1.77E-01

S-35 6.79E-01

4 (Nuclides obtained from 1977, 1979 worker bioassay lists) C-14

H-3

P-32

S-35

6, 16, 52 6: 184-in. Cyclotron (1957)

Advanced Light Source (present)

16: Sherwood Laboratory (1961)

Accelerator and Fusion Research (present)

52: General research (1961)

Accelerator and Fusion Research (present)

Radionuclides listed without a radionuclide fraction specified were 

obtained from a list of potential sources produced by accelerators 

(Building 6) identified in Patterson, Low-Beer, and Sargent (1969)c

U-238 2.90E-04

N-13 9.93E-01

O-15 5.19E-03

Ar-41 1.38E-03

Be-7

Co-57

Co-58

Co-60

Fe-55

Fe-59

Mn-54

Na-22

Na-24

Ni-57

Ni-63

Zn-65

10 Biomedical research (1986)

(nuclides obtained from 1977, 1979, 1982, 1983, 1986 worker 

bioassay lists)

C-14

Cr-51

H-3

P-32

S-35

Beta-gamma

15 (Nuclides obtained from 1977 worker bioassay lists) H-3 

25 Mechanical Technology

(nuclides obtained from 1977, 1979, 1982, 1983, 1986 worker 

bioassay lists)

Th

U-238

U

26, 76 Radioanalytical Laboratories (present) Ac-228 1.51E-08

Ag-108 2.65E-11

Am-241 4.15E-03

Am-243 2.56E-06

Ba-133 2.27E-04

Be-7 5.07E-10

Bk-249 9.91E-04

Br-76 6.94E-09

Br-77 3.78E-10

C-14 1.54E-02

Ce-144 5.72E-07

Cf-249 4.30E-04

Cf-250 7.57E-08

Cm-243 2.59E-07

Cm-244 2.12E-08

Cm-245 1.44E-10

Cm-246 1.13E-09
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 35 of 66

Current 

building no. Building activity and compounds Radionuclide Fraction

26, 76 

(Cont’d.)

Cm-248 4.54E-10

Co-56 9.83E-10

Co-57 7.00E-06

Co-58 5.29E-10

Co-60 1.64E-04

Cr-51 1.25E-10

Cs-134 9.63E-06

Cs-137 2.17E-03

Eu-152 3.93E-11

Fe-55 1.95E-05

H-3 1.77E-01

Hg-194 9.08E-09

I-125 3.80E-01

I-129 4.20E-05

I-131 3.81E-01

K-40 1.89E-09

Kr-76 1.51E-10

Mn-54 7.69E-06

Na-22 1.63E-08

Nb-95 1.63E-09

Ni-63 1.01E-05

Np-237 3.69E-06

Np-239 5.33E-08

Os-185 1.44E-09

P-32 3.78E-03

Pa-231 9.72E-08

Po-210 1.51E-08

Pu-238 5.46E-04

Pu-239 1.59E-03

Pu-241 3.40E-03

Pu-242 1.66E-08

Ra-226 2.27E-06

Ra-228 1.69E-04

Ru-106 2.27E-11

S-35 2.42E-02

Sb-125 8.81E-10

Sr-89 6.65E-07

Sr-90 1.38E-04

Tc-99 1.10E-06

Th-229 9.08E-08

Th-230 5.19E-06

Th-232 1.76E-05

U-232 3.79E-05

U-233 3.78E-03

U-235 4.17E-05

U-238 5.06E-04

Zn-65 1.20E-05

29 (Nuclides obtained from 1979 worker bioassay lists) Alpha

51 Bevatron

[Uranium and beta-gamma nuclides obtained from 1982, 1983, 

1986 worker bioassay lists; Am-241, Am-243, Bk-249, Cf-252, and 

Ra-226 obtained from LBNL (1995b); the remainder of the 

radionuclides were obtained from a list of potential sources 

produced by accelerators that were identified in Patterson, LowBeer, and Sargent (1969).c

] Building is currently undergoing 

decontamination and decommissioning.

U

Beta gamma

Am-241

Am-243

Bk-249

Cf-252

Ra-226

Be-7

Co-57

Co-58

Co-60

Fe-55

Fe-59

Mn-54

Na-22
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 36 of 66

Current 

building no. Building activity and compounds Radionuclide Fraction

51 (Cont’d.) Na-24

Ni-57

Ni-63

Zn-65

55, 56, 64 55: Animal house (1961)

Center for functional imaging and life sciences research (present)

56: Biomedical isotope facility (present)

64: Accelerator design (1961)

Life sciences research (present)

C-14 2.67E-08

Co-57 7.63E-08

Gd-153 2.29E-08

H-3 1.12E-05

I-123 4.40E-05

I-125 1.09E-04

I-131 6.02E-06

Nb-95 1.14E-07

P-32 4.58E-07

Ru-103 1.53E-07

Sn-113 7.63E-08

Tc-99m 3.16E-05

Tl-201 9.91E-07

F-18 1.00E+00

Alpha 3.71E-09

Beta 5.92E-08

57 (Nuclides obtained from 1982 worker bioassay lists) I-125

I-131

In-111

P-32

Sr-82

Sr-85

Sr-90

62 Materials and Molecular Research

(nuclides obtained from 1982, 1983, 1986 worker bioassay lists)

Th-228

70, 70A 70/70A: Nuclear Chemistry (1961)

70: Environmental energy technology, nuclear science, and earth 

sciences research (present)

70A: Nuclear, chemical, and life sciences research (present)

Ac-227d

Am-241 7.60E-06

Am-243 1.52E-04

Bi-207 7.60E-06

C-14 5.51E-01

Ca-45 1.20E-05

Ca-49 2.38E-08

Ce-141 2.59E-06

Cf-249 7.60E-07

Cf-252 7.60E-09

Cm-244 2.13E-08

Cm-248 1.08E-07

Co-60 3.61E-06

Cs-134 1.46E-06

Eu-152 1.07E-05

Fe-59 6.88E-03

H-3 2.76E-01

Hf-175 8.58E-06

Ho-166m 6.08E-06

I-125 2.25E-04

Mn-54 1.52E-07

Na-22 1.52E-06

Na-24 2.65E-05

Np-237 1.70E-03

P-32 4.13E-02

Pa-233 3.60E-06

Pu-238 4.35E-05

Pu-239 6.99E-06

Ra-226 1.52E-06

Rb-86 1.93E-04

Rh-101 7.60E-08

Ru-106c

S-35 1.06E-02

Sb-122 1.90E-07
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 37 of 66

Current 

building no. Building activity and compounds Radionuclide Fraction

70, 70A 

(Cont’d.)

Sb-124 1.10E-06

Sc-46 1.34E-03

Sc-49 1.19E-08

Sr-90 6.08E-06

Ta-182 1.52E-06

Tb-161c

Tc-99 8.04E-02

Th-229 2.17E-05

Th-232 9.44E-04

Tl-204 1.52E-06

Tm-170 1.52E-05

U-233 2.61E-04

U-234 1.08E-04

U-235 1.08E-04

U-238 3.81E-04

Y-90 1.22E-05

Yb-175 1.96E-06

Zr-88 4.56E-07

Zr-95 7.60E-04

Alpha 4.04E-03

Beta 2.38E-02

71 HILAC (1961)

Accelerator and fusion research (present)

Radionuclides listed without a radionuclide fraction specified were 

obtained from a list of potential sources produced by accelerators 

that were identified in Patterson, Low-Beer, and Sargent (1969)c

F-18 1.20E-08

H-3 1.00E+00

Be-7

Co-57

Co-58

Co-60

Fe-55

Fe-59

Mn-54

Na-22

Na-24

Ni-57

Ni-63

Zn-65

72 Health Physics (1961)

Low-background facility (present)

Ac-227 8.33E-08

Au-198 2.83E-06

Ba-133 2.99E-09

Bi-207 3.01E-08

Br-82 2.48E-07

Cd-113 6.40E-09

Co-58 9.23E-01

Co-60 1.20E-03

Cr-51 1.51E-06

Cu-64 4.96E-04

Eu-152 4.96E-04

Fe-59 2.95E-02

Hg-194 4.68E-08

Mn-54 2.68E-06

Na-24 9.92E-04

Nb-95 1.34E-09

P-32 9.50E-06

Pa-231 4.94E-08

Sb-122 2.97E-11

Sb-124 9.92E-04

Ta-182 5.28E-08

Th-228 5.23E-12

Th-229 2.62E-05

Th-232 4.96E-08

Ti-44 1.35E-07

Xe-133 3.97E-10

Zn-65 3.47E-11
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 38 of 66

Current 

building no. Building activity and compounds Radionuclide Fraction

72 (Cont’d.) Zr-95 4.10E-02

Zr-97 1.82E-03

74, 83, 84 74: Animal bioradiological laboratory (1961)

74, 83, 84: Life sciences research (present)

C-14 4.92E-03

H-3 5.05E-01

P-32 3.46E-02

S-35 4.56E-01

75 Radioisotope services (1961)

Former waste handling facility, operations ceased in 1997

Former National Tritium Labeling Facility operations ceased in 2001

H-3 1.00E+00

Alpha (Th-232) 1.69E-09

Beta (Sr-90) 5.72E-09

85 Hazardous waste handling facility (present) C-14 1.44E-01

H-3 8.56E-01

I-125 9.97E-06

Alpha (Th-232) 1.17E-06

Beta (Sr-90) 2.79E-06

88 88-in cyclotron (1961 – present)

Radionuclides listed without a radionuclide fraction specified were 

obtained from a list of potential sources produced by accelerators 

that were identified in Patterson, Low-Beer, and Sargent (1969)c

U-238 1.40E-10

C-11 9.94E-01

Cr-51 1.95E-12

Fe-55 8.19E-13

Fe-59 2.03E-12

Ge-71 6.37E-03

Zr-95 1.56E-12

Zr-97 2.34E-12

Y-88 3.12E-12

Alpha 4.72E-07

Beta 1.63E-06

Be-7

Co-57

Co-58

Co-60

Mn-54

Na-22

Na-24

Ni-57

Ni-63

Zn-65

934 Previously Leased Building use of biomedical radionuclides for DNA 

Labeling (nuclides obtained from 1982, 1983, 1986 worker bioassay 

lists)

C-14

H-3

I-125

P-32

S-35

977e H-3

P-32

P-33

S-35

C-14

Cd-109

a. Blank values in this table indicate not applicable or data not available. Unless otherwise noted, this table was derived from 

data reported for calendar years 2001 to 2004 as part of the National Emission Standards for Hazardous Air Pollutants 

(DOE 2002, 2003, 2004, 2005). Where stack emissions were not measured, LBNL derived a “receipts releasable” 

inventory of radionuclides available for release. While the emissions identified during this limited period are probably not

fully representative of the source term in each building, they do provide an estimate of the types of radionuclides to which 

employees might have been exposed. Activity fractions were obtained by dividing the activity total by the total activity for all 

radionuclides in a specific building/buildings over the 4 yr.

b. Buildings and radionuclides were added based on historical lists (e.g. bioassay lists, LBNL 2005a,b; Patterson, Low-Beer, 

and Sargent 1969). Radionuclide fractions were not determined for these.

c. Although Patterson, Low-Beer, and Sargent (1969) identified these radionuclides as potential exposure sources, the article 

concluded that “normal habits of cleanliness and occasionally the use of protective clothing ... are adequate to insure that 

the ordinary accelerator worker at LRL-Berkeley will not receive radiation exposure of any consequence from internal 

sources.”

d. Ac-227, Ru-106, and Tb-161 were added to the list of radionuclides for Building 70 because there were documented spills 

in Building 70 between 1957 and 1960 where these radionuclides were listed; no attempt to determine a radionuclide 

fraction has been made.

Additional table notes:

1. Radionuclides obtained from LBNL (2005a).

2. The information on the compounds was very limited and, therefore, this column has not been completed. A few of the 

WBCs reviewed contained some information about the compound, but most noted “unknown.”
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 39 of 66

6.0

This section describes the program for measuring skin and WB doses to workers from sources that 

were external to the body. Workers at LBNL were exposed to radiation from a variety of radioactive 

materials and radiation-producing machines. Personnel dosimeter records are generally available for 

all periods at LBNL for workers who had any potential for occupational radiation exposure. The 

operations and radiation safety staffs routinely reviewed dosimeter results for compliance with 

radiation control limits and investigated doses that approached annual or quarterly limits. LBNL used 

personnel dosimeters to measure and record doses from external radiation to designated workers 

throughout the history of its operations. These dosimeters included one or more of the following:

OCCUPATIONAL EXTERNAL DOSE

• Personnel WB beta/photon dosimeters 

• Pocket ionization chamber dosimeters (electrometers/electroscopes, etc.) 

• Personnel extremity dosimeters

• Personnel neutron dosimeters

LBNL began operations using dosimeter and processing technical support provided by the 

Metallurgical Laboratory at the University of Chicago. After 1952, LBNL operated its own fully 

functional personnel dosimetry program with in-house processing. Exposure data were recorded by 

film badges and thermoluminescent dosimeters (TLDs). Early exposure records also provided 

electroscope (or at times, electrometer or E was used) results, which supplemented the results 

measured by film. Dose reconstructors should use the electroscope results in a qualitative manner 

because no data were found on the calibration or energy response of these devices; they should use 

the film or TLD results to estimate the actual exposure or the electroscope reading if no corresponding 

dosimeter reading exists. The electroscope results include daily readings in tables captioned 

"dosimeter," "slow neutron," and "electroscope." The three readings, which occurred on the same 

dates, were evidently used to measure exposure at the end of work shifts [42]. There are data in 

worker records that show a comparison of the film dosimeter to electroscope records for the same 

interval; the data are similar and correlate, but they are not identical. This is probably due to a 

combination of differences in energy dependence, calibration techniques, where the dosimeters were 

worn in relation to each other, exposure geometry, etc.

Where the records bear the heading, "All readings are recorded in daily doses," the daily dose is 

100 mrem. The units are in fractions of 100 mrem (e.g., 0.08 = 8 mrem; Heinzelman 2003).

There are no records of reporting beta exposure using film from before 1979. This is presumed to be 

because the radiation protection staff at the time felt that deep dose was controlling to meet regulatory 

limits (i.e., that the shallow dose could not exceed the deep dose by a high enough amount to come 

close to the higher regulatory limit). The radiation protection staff was aware of the need to monitor 

for nonpenetrating radiation and used a film badge containing filters (LRL 1958e). Inasmuch as the 

limit for WB (deep) dose was 5 rem/yr and for skin (shallow) dose was 15 rem/yr, it is reasonable to 

assume that the shallow dose could not exceed the deep dose by more than a factor of 3 [43]. For 

cases where the shallow dose is important to reconstruction and is not reported, dose reconstructors 

should apply a factor of 3 to the WB dose to estimate the shallow dose.

Table 6-1 lists dosimeter types, periods of use, exchange frequencies, minimum detection limits 

(MDLs), and potential annual missed doses for LBNL. Tables 6-2 and 6-3 list the energies and 

percentages for neutron and beta/photon radiation, respectively. If more than one energy range and 

percentage combination is provided, dose reconstructors should use the combination that is most 

favorable to claimants. Table 6-4 summarizes recorded dose practices, and Table 6-5 summarizes 

the interpretation of the reported data. Table 6-6 lists adjustments to recorded dose, and Table 6-7 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 40 of 66

lists the bias and uncertainty of each specific dosimeter system. The adjustment to neutron dose by a 

factor of 2 provided in Table 6-6 should be used unless claimant records provide sufficient information 

to use the more specific ICRP Publication 60 correction factor provided in Table 6-3 (ICRP 1991). 

There is limited information about the use of extremity dosimeters before 1982. It appears that X-ray 

film was used as finger rings; there is no reference to wrist dosimeters (as were used at LANL or Oak 

Ridge National Laboratory) [44]. When extremity dose data is reported for a worker it should be 

assumed that the dosimeter characteristics were the same as for the general use dosimeters as 

described in Table 6-1. Estimates of missing dose can use dose results for coworkers or the recorded 

dose before and after the period of the missing dose. Unmonitored workers should be assigned 

environmental doses as discussed in Section 4 or coworker data, as appropriate. Assignment of dose 

to unmonitored workers is addressed in NIOSH (2002). 

Table 6-1. LBNL dosimeter type, period of use, exchange frequency, MDL, and potential annual 

missed dose. 

Period of use Dosimeter

MDLa 

(rem)

Exchange

frequencyb

Annual 

missed 

dosec (rem)

1941–1981d

[45]

Photon/electron - DuPont 558 and 519 filme 0.015 Weekly (n = 50) 0.375

0.020 Biweekly (n = 25) 0.25

0.030 Monthly (n = 12) 0.18

Neutron - Kodak NTA filmf 0.050 Weekly (n = 50) 1.25g

0.050 Biweekly (n = 25) 0.625 g

0.050 Monthly (n = 12) 0.30 g

1982–1985

[46]

Photon/electron/neutron - Harshaw TLD 

(TLD-100, TLD-200, TLD-600, and TLD-700)

0.010 Monthly (n = 12) 0.06

0.020 Quarterly (n = 4) 0.04

1985–present

[47]

Photon/electron - Panasonic 810AS and 802AS 

TLD

0.010 Monthly (n = 12) 0.06

0.015 Quarterly (n = 4) 0.03

0.025 Semiannual (n = 2) 0.025

Neutron - CR-39 0.010 Monthly (n = 12) 0.06

0.010 Quarterly (n = 4) 0.02

a. Estimated MDLs for each dosimeter technology. Dose values were recorded at levels less than the MDL.

b. Exchange frequencies were dependent on work assignment. If the exchange frequency is not evident based on trends 

in an individual’s personnel records, assume a monthly exchange frequency. If there is a possibility of more frequent 

monitoring than monthly that is not evident from review of worker data, the dose reconstructor should contact LBNL to 

request additional information.

c. Annual missed dose calculated using the MDL/2 method from NIOSH (2002).

d. From 1941 to 1952, ANL-E processed film dosimeters for LBNL. The procedure to process the dosimeter was 

described by Pardue, Goldstein, and Wollan (1944). From 1952 to 1955, LBNL processed film dosimeters using an 

equivalent method.

e. No record of the type of beta/gamma film used initially is available, but DuPont films were in use by ANL-E by the late 

1950s (ORAUT 2006).

f. The beginning of neutron monitoring is uncertain, but it appears to have been in use as early as 1951. Records indicate 

that films were developed but not routinely read before 1960. Films were apparently not evaluated unless there was a 

gamma dose measured for the same period (ORAUT 2006; CER 1981).

g. Neutron energies below 500 keV should be considered as unmonitored due to the insensitivity to neutrons with energies 

less than 500 keV (NIOSH 2002). Doses should be assigned in accordance with the guidelines in NIOSH (2002) and 

Table 6-6 below. 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 41 of 66

Table 6-2. Selection of beta and photon radiation energies and percentages [48]. 

Buildings Functional category

Radiation 

type

Energy 

selection (MeV) Percentage

1, 2, 3, 4, 5, 5A, 7, 8, 

9, 10, 11, 14, 16, 18, 

19, 20, 22, 24, 25, 

29, 38, 39, 55, 71

Chemistry: radioactive materials 

including Co-60, S-90, fission products, 

enriched uranium, depleted uranium, 

natural uranium, and others

Beta >15 100

Photon 30–250 

> 250

25

75

6, 26, 70, 70A, 88 Chemistry Heavy Elements Facility: 

Cf-252, Cm-244, Am-241, U-233, 

Pu-239, and others

Beta >15 100

Photon <30

30–250 

> 250

25

50

25

75 Waste Storage Yard: fission 

products, enriched uranium, natural 

uranium and others

Beta >15 100

Photon 30–250 

> 250

25

75

72, 74, 75 Irradiators: Co-60, Cs-137 Beta >15 100

Photon 30–250 

> 250

25

75

51, 71 Cyclotrons and Accelerators Beta >15 100

Photon <30 0

30-250 10

>250 90

Table 6-3. Selection of neutron radiation energies and percentages [49]. 

Buildings Description

Radiation 

type

Energy 

selection 

(MeV) Percentage

ICRP 60 

Correction 

Factor

1, 2, 3, 4, 5, 5A, 7, 

8, 9, 10, 11, 14, 16, 

18, 19, 20, 22, 24, 

25, 29, 38, 39, 55, 

71

Chemistry: radioactive materials 

including Co-60, S-90, fission 

products, enriched uranium, 

depleted uranium, natural 

uranium, and others

Neutron 0.1–2.0 100 1.91

6, 26, 70, 70A, 88 Chemistry Heavy Elements 

Facility: Cf-252 Cm-244, Am-241, 

U-233, Pu-239, and others

Neutron 0.1–2.0

2.0-20

90

10

1.71

0.13

75 Waste Storage Yard: fission 

products, enriched uranium, 

natural uranium, and others

Neutron 0.1–2.0

2.0-20

90

10

1.71

0.13

72, 74, 75 Irradiators: Co-60, Cs-137 Neutron 0.1–2.0 100 1.91

51, 71 Cyclotrons and Accelerators Neutron 0.1–2.0

2.0-20

50

50

0.95

0.65

Table 6-4. Recorded dose practices [50]. 

Period

Dosimeter measured 

quantities Compliance dose quantities

Photon/electron film dosimeter + NTA neutron dosimeter

1941–1981 Gamma (G)

Neutron (N)

WB = gamma + neutron

Photon/electron/neutron–Harshaw TLD

1982–1985 Gamma (G)

Neutron (N)

Total = photon + neutron (γ + η)

Skin = shallow = photon + neutron

Hand = finger extremity dose 

Photon/electron/neutron–Panasonic TLD + CR-39 neutron dosimeter

1985–present Skin

Photon 

Neutron 

Skin = photon + neutron (P/N)

WB = Photon + neutron
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 42 of 66

Table 6-5. Interpretation of reported data [51].

Period

Reported 

quantity Description

Interpretation 

of zeroes

Interpretation 

of blanks 

(no data)

Rollup of 

individual and 

annual data

Monitored/ 

unmonitored

1941–

1981

rem Reported WB doses 

include gamma. 

Neutron doses were 

designated with “N.” 

Claimant records 

indicate that some 

recorded results are 

expressed as a 

fraction of the daily 

dose limit of 100 

mrem.

Interpret reported zero 

as meaning less than 

MDL.

No data or blanks 

should be interpreted 

as individual was 

monitored with zero 

result.

Photon WB dose

Neutron WB dose

Total deep WB 

dose

From review of LBNL 

documents, it is not 

clear if all employees 

were continuously 

monitored. However, 

the monitoring records 

appear to be 

complete. It can be 

assumed if no 

monitoring records are 

included in an 

employee's file that

the individual was not 

monitored. 

1982–

1985

rem Reported WB doses 

qualified as either 

photon or neutron. 

Interpret reported zero 

as meaning less than 

MDL.

No data or blanks 

should be interpreted 

as individual was 

monitored with zero 

result.

Photon WB dose

Neutron WB dose

Shallow skin 

dose

Total deep WB 

dose

1985–

present

rem Photon deep, 

neutron deep, and 

skin dose reported.

Interpret reported zero 

as meaning less than 

MDL.

No data or blanks 

should be interpreted 

as individual was 

monitored with zero 

result.

Photon WB dose

Neutron WB dose

Shallow skin 

dose

Total deep WB 

dose

Table 6-6. Adjustments to recorded dose. 

Period Dosimeter Facility Adjustment to reported dose

1941–1985a Photon dosimetersb All facilities Use roentgen-to-organ dose conversion factors.

1985–present Photon dosimeters All facilities Use Hp(10)-to-organ dose conversion factors.

All years Neutron dosimetersc All facilities Multiply reported doses by factor of 2 to account for 

ICRP (1991) weighting factors.

a. Beta and nonpenetrating dose was not reported before 1982. In general, nonpenetrating radiation doses should be 

assigned as <30-keV photons if the employee worked with or around plutonium; otherwise, >15-keV electrons should be 

assigned (ORAUT 2005b). The guidance from Oak Ridge Associated Universities (ORAU) is:

If the nature of the nonpenetrating dose is unknown, consider the following guidance: 

1. For a likely noncompensable case, it is acceptable to assume the nonpenetrating dose is 

associated with <30-keV photons, as this maximizes POC. 

2. For a likely compensable case, it is acceptable to assume the nonpenetrating dose is 

associated with >15-keV electrons, as this minimizes POC.

3. If the compensability decision may hinge on this issue, and if the partitioning of the 

nonpenetrating dose cannot be decided based on the available information, additional research 

might be required.

b. The doses from low-energy photons (<30 keV) were underestimated for 1982 to 1985 as a result of the design of the 

dosimeter holder (ORAUT 2005c). The dose from <30-keV photons should be calculated by the following formula:

Photon 30kev =< (shallow WB ×− 2)

c. The NTA dosimeter (1941 to 1969) exhibited a lower energy threshold of approximately 500 keV (NIOSH 2002). The 

photon dose was measured adequately and all LBNL neutron dose was accompanied by a significant photon dose. For 

neutron dose received before 1969, the dose should be adjusted by using a neutron-to-photon ratio. The neutron 

exposure was determined with a neutron-to-photon ratio of 0.73 ±2.1 [52]. The geometric mean was 0.73, and the 

geometric standard deviation was 2.1[53]. The assumed upper 95th percentile was assumed to be 2.47 [54]. 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 43 of 66

Table 6-7. Bias and uncertainty [55]. 

Site-specific dosimetry system

Bias magnitude and range Uncertainty factors

Overall biasa Range in biasb Systematicc Randomd

Photon/electron film (1941–1981) nonplutonium

facility (as listed in Tables 2-1 and 2-2)

1.27 1.23–1.60 1.2 1.8

Neutron NTA film (1941–1981) nonplutonium

facility (as listed in Tables 2-1 and 2-2)

1.0 0.5–2.0 1.05 2.0

Photon/electron film (1941–1981) plutonium

facility (as listed in Tables 2-1 and 2-2)

1.0 0.25–2.0 1.05 1.3

Neutron NTA film (1941–1981) plutonium

facility (as listed in Tables 2-1 and 2-2)

1.0 0.5–2.0 1.05 2.0

Harshaw photon/electron/neutron TLD (1982–

1985)

1.12 1.04–1.2 1.05 1.2

Panasonic photon/electron TLD (1985–present) 1.0 0.8–1.2 1.05 1.2

CR-39 neutron (1985–present) 1.0 0.6–1.5 1.05 1.5

a. Based on the distribution of energy levels and geometry judged most likely. Divide recorded dose by the table's bias 

value to calculate deep dose.

b. Range of overall bias factors based on alternative distributions of energy levels and geometry.

c. Systematic uncertainty resulting from lack of knowledge about actual distributions of energy levels and geometry.

d. Random uncertainty resulting from variation among workers in energy levels and geometry.

7.0 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.

[1] Turpin, Baynard. ORAU Team, Task 3. May, 2006. “About Berkeley Lab” on the Lawrence 

Berkeley National Laboratory home page found at: http://www.lbl.gov/LBL-PID/LBLOverview.html. 

[2] Turpin, Baynard. ORAU Team, Task 3. May, 2006. Table 2-1 information was 

obtained from a document titled: RadMap, Legacy and Current Radioactive Material 

Locations, by Michael R. Dupray of LBNL.

[3] Turpin, Baynard. ORAU Team, Task 3. May, 2006. 1995 LBNL Site Environmental Report, 

Introduction and Purpose, found at: 

http://www.lbl.gov/ehs/esg/95ser/95serchap1.htm#Introduction%20and%20Purpose

[4] Turpin, Baynard. ORAU Team, Task 3. May, 2006. “About Berkeley Lab” on the Lawrence 

Berkeley National Laboratory home page found at: http://www.lbl.gov/LBL-PID/LBLOverview.html

[5] Turpin, Baynard. ORAU Team, Task 3. May, 2006, Table 2-2 information was obtained from 

the same source as item 2-2 and Project Document number 030007011, “Radioisotope 

Inventory, LRL-Berkeley “Appendix B Spaces,” March 17, 1964. [SRDB Ref ID: 20693]

[6] De Castro, T. M., and S. B. Thomas, LBNL, Health Physics Department, 1975.
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 44 of 66

That the collimation and filtration were checked in the survey of March 19, 1975 and found to 

be acceptable is a relevant indication that the equipment was operating properly.

[7] ORAUT-OTIB-0006, Dose Reconstruction from Occupationally Related Diagnostic X-Ray 

Procedures.

Distances selected for computations are consistent with this document.

[8] Bradfield, N. E., LBNL, X-Ray System Supervisor, 1987.

This document has numerous handwritten notes, ending on last page with “Changes made 

9/28/87 Installation of New Picker”, signed by Nancy E. Bradfield, RN. Document establishes 

date of replacement for X-ray equipment.

[9] Turner, James E. ORAU Team, Task 3. May, 2006.

The assumption of an annual X-ray screening frequency is consistent with common practice at 

other laboratories. Cases with reconstructed doses in the NOCTS files for LBNL have used an 

assumed annual frequency.

[10] East, James. ORAU Team, Task 3. May, 2006. Summary statement to provide general 

background on the quality and effectiveness of the monitoring program. Based on Subject 

Expert’s perception of referenced documents.

[11] East, James. ORAU Team, Task 3. May, 2006. Various papers and memoranda in the early 

50’s were presented on the subject regarding the design and effectiveness. Cited 1950 

memorandum on hood design for radiochemistry specifying many features still in use today. 

Background information.

[12] East, James. ORAU Team, Task 3. May, 2006. Summary of the environmental monitoring 

data, from the beginning of observations in 1959 show a significant drop in ambient dose 

rates. This is perceived as a result of a combination of factors, reduction in the weapons 

testing fallout, improved instrumentation and monitoring program, and implementation of the 

ALARA concept. Programs evolved from portable instrument spot checks to use of continuous 

monitoring with film badges and later TLDs.

[13] East, James. ORAU Team, Task 3. May, 2006. Many years only reported maximum 

observed values. Later, average values were also reported. For consistency, and to be 

favorable to claimants, the maximum value was chosen for all years.

[14] East, James. ORAU Team, Task 3. May, 2006. No one worker received the maximum value 

every year, and the requirements of the computer code for inputs of average and maximum 

uncertainty lead to the estimates provided and remain favorable to claimants.

[15] East, James. ORAU Team, Task 3. May, 2006. Summary statement based on facts reported 

in the Environmental Reports over the years. The portable instrument readings were 

converted from hourly rates to annual for consistency.

[16] East, James. ORAU Team, Task 3. May, 2006. This statement is drawn from notations in 

many of the Environmental Reports and represents the capabilities of the monitoring at the 

time observations were made.
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 45 of 66

[17] East, James. ORAU Team, Task 3. May, 2006. Based on data presented in available 

environmental reports listed in the Reference section. Background information to support the 

estimates provided in the next paragraph.

[18] East, James. ORAU Team, Task 3. May, 2006. The growth in the power of the accelerators 

is viewed to be almost exponential. The back side of the curve (1959 on) shows a sharp drop 

in reported doses. A linear estimate of these years would overestimate observed doses. 

These two observations provide logical basis for the statement.

[19] East, James. ORAU Team, Task 3. May, 2006. Statement of the obvious effects of applying 

average dose observed over the early years, overestimating earliest years, and 

underestimating the later.

[20] East, James. ORAU Team, Task 3. May, 2006. The parabolic curve best parallels the growth 

in accelerator power, and mirrors the decline in observed doses in the following years.

[21] East, James. ORAU Team, Task 3. May, 2006. The previous paragraphs provide the reason 

for selecting the parabolic curve, and this paragraph summarizes the effects of doing so.

[22] East, James. ORAU Team, Task 3. May, 2006. The internal dose calculated in the record by 

LBNL are from releases and based on Th for alpha and Sr for beta because these were 

predominate and/or have the highest dose per activity factors and are favorable to the 

claimant.

[23] East, James. ORAU Team, Task 3. May, 2006. As stated, the average value for the 

observed years provides a good estimate of the other years. There appears to be little or no 

impact from LBNL operations, nor should there be from accelerator operations.

[24] East, James. ORAU Team, Task 3. May, 2006. Some of the recent environmental reports 

did not include beta activity in air, necessitating the estimates for the missing years. These

estimates were based on recent years that were reported, and a round number above this 

trend was selected to be favorable to the claimant.

[25] East, James. ORAU Team, Task 3. May, 2006. Justification for the estimates is provided in 

the text for the years where observations are not available. Multiple effects are being 

addressed, the rise in airborne concentrations from atmospheric weapons testing and the 

increasing accelerator power and resulting generation of beta airborne concentrations, most of 

which are short lived and/or have low dose to activity conversion factors.

[26] East, James. ORAU Team, Task 3. May, 2006. The measurement for H-3 in 2004 was not 

reported. An estimate of H-3 was based on a H-3 to C-14 ratio. The ratio chosen is 

representative (average) of the data and favorable to claimant. Further discussion of similar 

technique is provided in the paragraph on C-14.

[27] East, James. ORAU Team, Task 3. May, 2006. Measurements were not available prior to 

1974. It is well established that H-3 and C-14 are common byproducts of accelerator 

operation. It is reasonable to assume that the rise in airborne concentrations is proportional to 

accelerator power and operation time. The linear estimate from the peak downwards towards 

initial operations provides for concentrations favorable to the claimant. This upward rise was 

controlled in the later years with aggressive shielding design with steps in power level, see 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 46 of 66

environmental reports for summaries of the shielding efforts for the new and modified 

accelerator projects.

[28] East, James. ORAU Team, Task 3. May, 2006. The ratio of C-14 to H-3 was calculated for 

each of the years where both values were reported (1974-1995). The average of the twentytwo observations was 0.155, the maximum was 0.73 and minimum of 0.02. This factor was 

applied to C-14 from 1996 to 2003 (except for 1999 when a value of 2592 was reported. As 

discussed above, the only year this was applied to H-3 in reverse was 2004, the only year that 

C-14 was reported and H-3 was not reported.

[29] Langille, Elizabeth A. ORAU Team, Task 3. May, 2006. Information on the decision to 

eliminate routine bioassays and use operational bioassays (based on scheduled work) was 

obtained during a telephone conversation between Elizabeth Langille and James Floyd (LBL 

Group Leader EH&S), May 6,2006 [SRDB Ref. ID 23402]. 

Documentation for the approach is also found in:

AIMahamid, I., 2005, Technical Basis Document for Internal Dosimetry, EH&S Procedure 301, 

Rev.4, University of California, Lawrence Berkeley National Laboratory, Berkeley, California, 

May 12. [SRDB Ref. ID: 21162] 

[30] Langille, Elizabeth A. ORAU Team, Task 3. May, 2006. “LBNL Electronic Record of Results 

1960 – 1996,” ORAU Team 2006. [SRDB Ref. ID: 23841] 

The electronic record of results was obtained from J. Floyd, LBNL Group Leader EH&S and 

includes a listing of method codes for the bioassay sample analysis results table. Where 

specific analyses were found in NOCTS records or in other SRDB records and could be 

compared to the electronic records, the method codes were verified to match those in the 

electronic records. 

[31] Sargent, Thornton III, 1991, LBNL, Whole Body Counter Facility Operations Manual, January 

22. [SRDB Ref. ID: 20900]

The operator aid specifies that a 15 minute count time is used but then provides a step to 

“assess the need for other counts in other geometries, as per detailed instructions.” These 

instructions implied that special instructions might include an increased count time to achieve 

a desired detection limit.

[32] Low-Beer, A. DeG., 1963, “Report of Bioassay Program for 1962,” memorandum to Dr. H. G. 

Parker, University of California, Lawrence Radiation Laboratory, Berkeley, California, January 

31. [SRDB Ref. ID: 21252] 

The analysis of special fecal samples was noted in this document for 1962. Earlier special 

fecal samples are also noted in the electronic record of results found in SRDB Ref. ID 23481.

[33] Low-Beer, A. DeG., 1963, “Report of Bioassay Program for 1962,” memorandum to Dr. H. G. 

Parker, University of California, Lawrence Radiation Laboratory, Berkeley, California, January 

31. [SRDB Ref. ID: 21252]

The analysis of special blood samples was noted in this document for 1962. Earlier special 

blood (SB) samples are also noted in the electronic record of results found in SRDB Ref. ID 

23481.

[34] LBNL, 1983, Radiochemical Assay Report, July 12. [SRDB Ref. ID: 20880]

The record showed that spot urine collection was used for H-3 analysis during this time period. 

Spot urine collection may also have been used for H-3 earlier than 1983 at LBNL but of the 

records reviewed, this was the earliest date noted.
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 47 of 66

[35] LBNL (Earnest Orlando Lawrence Berkeley National Laboratory), 1995, Technical Basis 

Document for Internal Dosimetry, University of California, Berkeley, California, May 12. [SRDB 

Ref. ID: 20926]

This internal dosimetry program document provides information that routine 24 hour urine 

bioassay samples were obtained and analyzed for certain workers. 

[36] LBNL (Earnest Orlando Lawrence Berkeley National Laboratory), 1995, Technical Basis 

Document for Internal Dosimetry, University of California, Berkeley, California, May 12. [SRDB 

Ref. ID: 20926]

This internal dosimetry program document provides information that fecal bioassay samples 

were obtained and analyzed as special bioassays when determined necessary.

[37] LBNL (Earnest Orlando Lawrence Berkeley National Laboratory), 1995, Technical Basis 

Document for Internal Dosimetry, University of California, Berkeley, California, May 12. [SRDB 

Ref. ID: 20926]

This internal dosimetry program document provides information that spot urine bioassay 

samples were obtained and analyzed as determined necessary by EH&S.

[38] LBNL (Ernest Orlando Lawrence Berkeley National Laboratory), 1995, Technical Basis 

Document for Internal Dosimetry, University of California, Berkeley, California, May 12. 

[SRDB Ref. ID: 20926]

The in vivo method and detection limits found in the table were taken from this technical basis 

document.

[39] LBNL (Ernest Orlando Lawrence Berkeley National Laboratory), 1995, Technical Basis 

Document for Internal Dosimetry, University of California, Berkeley, California, May 12. 

[SRDB Ref. ID: 20926]

The in vivo method and detection limits found in the table were taken from this technical basis 

document.

[40] Langille, Elizabeth A. ORAU Team, Task 3. May 31, 2006. A review of the bioassay records 

in approximately 100 claims on NOCTS showed a file in 1971 (Claim 16237) where the 

individual was counted on a WBC and an analysis for P-32 was performed. This file was used 

to establish the beginning time for this method as 1971. Using WBC to analyze for P-32 was 

also noted in 1995 LBNL Technical Basis document for internal dosimetry and so the method 

was assumed to be available until LBNL stopped performing WBC in 1996.

LBNL (Ernest Orlando Lawrence Berkeley National Laboratory), 1995, Technical Basis 

Document for Internal Dosimetry, University of California, Berkeley, California, May 12. 

[SRDB Ref. ID: 20926]

The in vivo method and detection limits found in the table were taken from this technical basis 

document.

[41] Langille, Elizabeth A. ORAU Team, Task 3. May, 2006. “LBNL Electronic Record of Results 

1960 – 1996,” ORAU Team 2006. [SRDB Ref. ID: 23841] 

The electronic record of results was obtained from J. Floyd, LBNL Group Leader EH&S and 

includes a listing of method codes for the bioassay sample analysis results table. Where 

specific analyses were found in NOCTS records or in other SRDB records and could be 

compared to the electronic records, the method codes were verified to match those in the 

electronic records. 
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 48 of 66

[42] Guido, Joseph, ORAU Team, Task 5. June 5, 2006. Mr. Guido reviewed the original data 

from multiple claims and made the observation that the readings represented the end of work 

shifts. 

[43] Thomas, Bill R. ORAU Team, Task 3. June 5, 2006. Mr. Thomas reviewed the manual for 

radiation safety practices published in 1958 (LRL, 1958e). It appeared that the RADCON staff 

implemented a procedure that limited the deep dose and adequately limited the shallow dose. 

In the absence of other data, a ratio was established for shallow exposure to deep exposure of 

3:1.

[44] Guido, Joseph. ORAU Team, Task 5. June 5, 2006. Mr. Guido and Mr. Thomas discussed 

the lack of results for extremity dose prior to 1982. Mr. Guido summarized the records from 

several claimants. Wrist dosimeters were used at LANL and ORNL for the same time period; 

it appeared that the technology was available. 

[45] Smith, Matthew H. ORAU Team. May 22, 2006. Mr. Smith and Mr. Thomas discussed the 

limited response of NTA film to neutrons with an energy below 800 keV. There were no 

documented studies of the poor response provided by LBNL. Mr. Smith suggested a response 

similar to that selected in the Hanford TBD (ORAUT, 2004) and information derived for the 

Rocky Flats Plant in OTIB-0027 (ORAUT, 2005c). It was necessary to establish a neutron to 

photon ratio in order to modify the NTA dosimeter results. For operations handling 

transuranics including plutonium, the neutron to photon ratio was judged to be similar to 

Hanford plutonium operations with a geometric mean of 0.73, GSD 2.1, and upper 95% 2.47 

(ORAUT, 2004). 

[46] Smith, Matthew H. ORAU Team. May 22, 2006. Mr. Smith and Mr. Thomas discussed the 

limited response of NTA film to neutrons with an energy below 800 kev. There were no 

documented studies of the poor response provided by LBNL. Mr. Smith suggested a response 

similar to that selected in the Hanford TBD (ORAUT, 2004) and information derived for the 

Rocky Flats Plant in OTIB-0027 (ORAUT, 2005c). It was necessary to establish a neutron to 

photon ratio in order to modify the NTA dosimeter results. For operations handling 

transuranics including plutonium, the neutron to photon ratio was judged to be similar to 

Hanford plutonium operations with a geometric mean of 0.73, GSD 2.1, and upper 95% 2.47 

(ORAUT, 2004). 

[47] Smith, Matthew H. ORAU Team. May 22, 2006. Mr. Smith and Mr. Thomas discussed the 

limited response of NTA film to neutrons with an energy below 800 kev. There were no 

documented studies of the poor response provided by LBNL. Mr. Smith suggested a response 

similar to that selected in the Hanford TBD (ORAUT, 2004) and information derived for the 

Rocky Flats Plant in OTIB-0027 (ORAUT, 2005c). It was necessary to establish a neutron to 

photon ratio in order to modify the NTA dosimeter results. For operations handling 

transuranics including plutonium, the neutron to photon ratio was judged to be similar to 

Hanford plutonium operations with a geometric mean of 0.73, GSD 2.1, and upper 95% 2.47 

(ORAUT, 2004). 

[48] Thomas, Bill R. ORAU Team, Task 3. May 1, 2006. Mr Thomas reviewed the types of 

activities described in Chapter 2 of the LBNL TBD. Given the type of operations, the energy 

fractions were selected in a manner similar to the LLNL and LANL facilities.
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 49 of 66

[49] Thomas, Bill R. ORAU Team, Task 3. May 1, 2006. Mr Thomas reviewed the types of 

activities described in Chapter 2 of the LBNL TBD. Given the type of operations, the energy 

fractions were selected in a manner similar to the LLNL and LANL facilities.

[50] Thomas, Bill R. ORAU Team, Task 3. May 1, 2006. Mr. Thomas reviewed the original data 

from multiple claims in the NOCTS database and made the observation that codes assigned to 

external dosimetry were updated overtime. The types of dosimeters were described in the 

documentation provided with the exposure data.

[51] Thomas, Bill R. ORAU Team, Task 3. May 1, 2006. Mr. Thomas reviewed the original data 

from multiple claims in the NOCTS database and made the observation that codes assigned to 

external dosimetry were updated over time. The types of dosimeters were described in the 

documentation provided with the exposure data.

[52] Thomas, Bill R. ORAU Team, Task 3. May 1, 2006. Mr. Thomas reviewed the results 

provided in the NOCTS database to identify the frequency that dosimeters were exchanged. 

Design and type of film was described in the Argonne East TBD (ORAUT, 2006). Mr. Joe 

Guido, ORAU Task 5, June 5, 2006 confirmed that film badges were used through 1981 by 

reviewing claims in the NOCTS database. 

[53] Thomas, Bill R. ORAU Team, Task 3. May 1, 2006. Mr. Thomas reviewed the results 

provided in the NOCTS database to identify the frequency that dosimeters were exchanged. 

The MDL was equivalent to that of the LLNL and LANL programs for similar dosimeters and 

processing procedures.

[54] Thomas, Bill R. ORAU Team, Task 3. May 1, 2006. Mr. Thomas reviewed the results 

provided in the NOCTS database to identify the frequency that dosimeters were exchanged. 

The MDL was equivalent to that of the LLNL and LANL programs.

[55] Smith, Matthew H. ORAU Team. May 22, 2006. Mr Smith reviewed that bias and uncertainty 

described for the types of dosimeters and LBNL facilities. He recommended that the table be 

revised to be consistent with the Hanford TBD (ORAUT, 2004) and information derived for the 

Rocky Flats Plant in OTIB-0027 (ORAUT, 2005C).
Document No. ORAUT-TKBS-0049 Revision No. 01 Effective Date: 04/02/2007 Page 50 of 66

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GLOSSARY

absorbed dose

Amount of energy in rads or grays deposited in a substance by ionizing radiation per unit mass 

of the substance. See dose. 

absorption type 

Categories for materials according to their speed of absorption in the body, which replaced the 

earlier inhalation clearance classes. Defined by the International Commission on Radiological 

Protection, the absorption types are F for fast absorption (formerly inhalation class D), M for 

moderate absorption (formerly inhalation class W), and S for slow absorption (formerly 

inhalation class Y). Also called solubility type. 

activity median aerodynamic diameter (AMAD)

Diameter of a unit density sphere with the same terminal settling velocity in air as that of the 

aerosol particle whose activity is the median for the entire aerosol. In relation to health 

physics, normally assumed to be 5 micrometers. Also called aerodynamic equivalent 

diameter. 

acute exposure

Radiation exposure to the body delivered in a short period. See chronic exposure. 

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) becquerels. 

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.
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bioassay

Determination of kinds, quantities, or concentrations, and in some cases locations of 

radioactive material in a living body, whether by direct measurement (in vivo measurement) or 

by analysis and evaluation of materials excreted or removed from the body (in vitro

measurement). Also called radiobioassay. 

body burden

Amount of radioactive material in an individual's body at a particular point in time.

chronic exposure 

Radiation dose to the body delivered in small amounts over a long period (e.g., days or years). 

For the purposes of input to the Interactive RadioEpidemiological Program, chronic exposure 

is selected under exposure rate for beta and neutron dose. See acute exposure. 

curie (Ci) 

Traditional unit of radioactivity equal to 37 billion (3.7 x 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

depleted uranium (DU) 

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium. As 

examples, spent (used) fuel elements, byproduct tails, residues from uranium isotope 

separation, and some weapons materials contain DU. DU can be blended with highly 

enriched uranium to make reactor fuel or used as a raw material to produce plutonium.

dose 

In general, the effects of ionizing radiation in terms of the specific amount of energy absorbed 

per unit of mass. Effective and equivalent doses are in units of rem or sievert; other types of 

dose are in units of roentgens, rads, reps, or grays. Various terms narrow the type of dose, 

and some are additive: 

• Absorbed dose is the amount of energy deposited in a substance by ionizing radiation.

• Collective dose is the sum of the doses to a specific population.

• Committed dose is the dose over time (usually 50 years for workers) to a specific organ 

or tissue from an intake of radioactive material.

• Cumulative dose is the sum of all doses to the same portion of the body or to the whole 

body over time.

• Deep dose is the dose at a 1-centimeter depth in tissue (1,000 milligrams per square 

centimeter).

• Effective dose is the sum of the equivalent doses in the principal tissues and organs of

the body, each weighted by a tissue weighting factor that accounts for the probabilities 

of fatal and nonfatal cancers according to severity and the average length of life lost 

due to an induced cancer. It indicates the biological effect of the radiation exposure in 

that tissue. 

• Equivalent dose or dose equivalent is the absorbed dose in a tissue or organ multiplied 

by a weighting factor for the particular type of radiation. 
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• Organ dose is the dose to a specific organ.

• Penetrating dose is that from higher energy photon (gamma and X-ray) radiation and 

neutron radiation that penetrates the outer layers of the skin. Nonpenetrating dose is 

that from beta and lower energy photon radiation. 

• Personal dose equivalent is the dose equivalent in soft tissue below a specified point 

on the body at a specified depth. 

• Shallow dose is the dose at a 0.07-millimeter depth in tissue (7 milligrams per square 

centimeter).

• Skin dose is the dose to the skin.

• Whole-body dose is the dose to the entire body excluding the contents of the 

gastrointestinal tract, urinary bladder, and gall bladder.

deep dose equivalent (DDE, Hd, Hp(10)) 

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

per square centimeter). See dose. 

depleted uranium (DU)

Uranium with a percentage of 235U lower than the 0.7% found in natural uranium. As 

examples, spent (used) fuel elements, byproduct tails, residues from uranium isotope 

separation, and some weapons materials contain DU. DU can be blended with highly 

enriched uranium to make reactor fuel or used as a raw material to produce plutonium.

dose equivalent (H, DE) 

Product of absorbed dose in units of rem or sievert in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose. 

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See film dosimeter, neutron film dosimeter, and thermoluminescent

dosimeter. 

dosimetry

Measurement and calculation of internal and external radiation doses.

dosimetry system

System for assessment of received radiation dose. This includes the fabrication, assignment, 

and processing of external dosimeters, and/or the collection and analysis of bioassay samples, 

and the interpretation and documentation of the results.

enriched uranium (EU)

Uranium in which processing has increased the proportion of 235U to 238U to above the natural 

level of 0.7%. Reactor-grade uranium is usually about 3.5% 235U; weapons-grade uranium 

contains greater than 90% 235U.
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exposure

In general, the act of being exposed to ionizing radiation. See acute exposure and chronic 

exposure. In this document, exposure does not refer to the radiological physics concept of 

charge liberated per unit mass of air.

film 

Radiation-sensitive photographic film in a light-tight wrapping. See film dosimeter. 

film dosimeter

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge. 

fission products

(1) Radionuclides produced by fission or by the subsequent radioactive decay of 

radionuclides. (2) Fragments other than neutrons that result from the splitting of an atomic 

nucleus. 

gamma radiation 

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma rays are very 

penetrating, but dense materials such as lead or uranium or thick structures can stop them. 

Gamma photons are identical to X-ray photons of high energy; the difference is that X-rays do 

not originate in the nucleus. 

gray (Gy) 

International System unit of absorbed radiation dose, which is the amount of energy from any 

type of ionizing radiation deposited in any medium; 1 Gy equals 1 joule per kilogram or 

100 rads.

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons.

intake

Radioactive material taken into the body by inhalation, absorption through the skin, injection, 

ingestion, or through wounds. Depending on the radionuclide involved, intakes are in units of 

mass, activity, or potential alpha energy.

internal dose or exposure

Dose received from radioactive material in the body.

internal dose assessment

Estimation of an intake of radioactive material and the consequent internal radiation dose 

based on measurements in the work environment and/or bioassay. 
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in vitro

Of or relating to a process that takes place under artificial conditions or outside a living 

organism (e.g., in the laboratory). From Latin meaning in glass. 

in vivo

Of or relating to a process that takes place in a living organism. From Latin meaning in life. 

lung solubility type

See absorption type. 

minimum detectable activity or amount (MDA)

Lowest amount of radioactive activity or substance amount detectable by a specific instrument 

or process. Smallest amount or activity of a radionuclide in a sample or organ that yields a 

result above the detection level with a specific probability of a Type II (false negative) error 

while accepting an specific probability of a Type I (false positive) error.

minimum detection level (MDL)

Lowest amount (mass or activity) of a substance detectable by a specific instrument or 

process. Often assumed to be the level at which a dose is detected at the 2-sigma level (i.e., 

95% of the time). Also called minimum detectable limit and minimum detection limit or level. 

missed dose

Dose to monitored workers that was not measured or recorded due to such factors as a 

missing or damaged dosimeter or a result below the detection limits of the dosimeter. Missed 

dose is especially important in the early years of radiation monitoring, when relatively high 

detection limits were combined with short exchange periods. 

monitoring

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, ground water, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 

monitoring includes internal and external dosimetry including interpretation of the 

measurements.

natural uranium (U, U-nat, NU) 

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

weight. The specific activity of this mixture is 2.6 × 107 becquerels per kilogram 

(0.7 picocuries per gram). See uranium. 

neutron

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

neutron film dosimeter

Film dosimeter with a nuclear track emulsion, type A, film packet.

nuclear track emulsion, Type A (NTA)

Film sensitive to fast neutrons made by Eastman Kodak. The developed image has tracks 

caused by neutrons that visible under oil immersion with about 1,000-power magnification.
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occupational dose

Internal and external ionizing radiation dose from exposure during employment. Occupational 

dose does not include that from background radiation or medical diagnostics, research, or 

treatment.

occupational exposure

Exposure to radiation and/or to radioactive material from sources of radiation, whether in the 

possession of the DOE site contractor or other person, in a restricted area or in the course of 

employment in which the individual's assigned duties. Occupational exposure does not 

include exposure to background radiation, as a patient from medical practices, from voluntary 

participation in medical research programs, or as a member of the public.

occupational medical exposure

Exposure to radiation or radioactive materials from medical diagnostic procedures during 

physical examinations that are required as a condition of employment. For dose 

reconstruction, occupational medical exposure is a component of occupational exposure.

personal dose equivalent [Hp(d)] 

Dose equivalent in units of rem or sievert in soft tissue below a specified point on the body at 

an appropriate depth d. The depths selected for personal dosimetry are 0.07 millimeters 

(7 milligrams per square centimeter) and 10 millimeters (1,000 milligrams per square 

centimeter), respectively, for the skin (shallow) and whole-body (deep) doses. These are 

noted as Hp(0.07) and Hp(10), respectively. The International Commission on Radiological 

Measurement and Units recommended Hp(d) in 1993 as dose quantity for radiological 

protection.

photon

Basic unit of electromagnetic radiation. Photons are massless “packages” of light energy that 

range from low-energy microwave photons to high-energy gamma rays. Photons have 

energies between 10 and 100 kiloelectron-volts. See photon radiation. 

photon radiation

Electromagnetic radiation of light energy (photons) from microwaves to gamma rays. Gamma 

rays and X-rays are examples of ionizing photon radiation, which have enough energy to 

penetrate matter, including the body, and deposit energy in that matter.

probability of causation (POC)

For dose reconstruction under the Energy Employees Occupational Illness Compensation Act, 

the percent likelihood that a worker incurred a particular cancer from occupational exposure to 

radiation.

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joule per kilogram) of absorbing tissue. The rad has 

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose. 
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radiation

Subatomic particles and electromagnetic rays (photons) that travel from one point to another, 

some of which can pass through or partly through solid materials including the human body. 

See ionizing radiation. 

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.

radiograph

Photographic image produced on film by gamma rays or X-rays. Some of the rays (photons) 

can pass through parts of an item, while more opaque parts partially or completely absorb 

them and thus cast a shadow on the film.

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.

roentgen (R)

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulomb per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0°C and standard atmospheric pressure. An 

exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for 

higher energy photons (generally greater than 100 kiloelectron-volts).

routine monitoring

Monitoring carried out at regular intervals during normal operations. See special monitoring. 

shallow absorbed dose (Ds) 

Absorbed dose at a depth of 0.07 centimeters (7 milligrams per square centimeter) in a 

material of specified geometry and composition.

shallow dose equivalent [SDE, Hs, Hp(0.07)] 

Dose equivalent in units of rem or sievert at a depth of 0.07 millimeters (7 milligrams per 

square centimeter) in tissue equal to the sum of the penetrating and nonpenetrating doses.

sievert (Sv) 

International System unit for dose equivalent, which indicates the biological damage caused 

by radiation. The unit is the radiation value in gray (equal to 1 joule per kilogram) multiplied by 

a weighting factor for the type of radiation and a weighting factor for the tissue; 1 Sv equals 

100 rem.

special monitoring

Monitoring in addition to the routine monitoring program carried out for special purposes such 

as estimating the amount of radionuclide deposited in a person after a known or suspected 

accidental intake or after a known or suspected environmental release. See routine 

monitoring. 
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spot sample

In relation to bioassay, usually a single void of urine.

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. 

transuranic (TRU) elements 

Elements with atomic numbers above 92 (uranium). Examples include plutonium and 

americium. All isotopes of the transuranic elements are radioactive, they are naturally either 

rare or nonexistent on Earth, and most are known only as a result of research using nuclear 

reactors and particle accelerators because of extremely short half-lives.

unmonitored dose

Potential unrecorded dose that could have resulted because a worker was not monitored. See 

missed dose. 

uranium (U)

Heavy, metallic, and radioactive element with atomic number 92. Most natural uranium as 

found in ores is 238U with trace levels of other isotopes. Uranium-235 (0.7% of natural 

uranium) is fissile by itself and used in nuclear weapons as well as reactors. Uranium-238 

(99.3% of natural uranium) is fissionable by fast neutrons and used in nuclear reactors. 

Natural uranium contains a minute amount of 234U. See depleted uranium, enriched uranium, 

and natural uranium. 

whole-body (WB) dose 

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 

10 millimeters (1,000 milligrams per square centimeter). Also called penetrating dose. 

See dose. 

X-ray 

(1) See X-ray radiation. (2) See radiograph. 

X-ray radiation

Penetrating electromagnetic radiation (photons) of short wavelength (0.001 to 10 nanometers) 

and energy less than 250 kiloelectron-volts. X-rays usually come from excitation of the 

electron field around certain nuclei. Once formed, there is no difference between X-rays and 

gamma rays, but gamma photons originate inside the nucleus of an atom.