ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 115

DOE Review Release 05/03/2010

Document Title:

Site Profile for the Brookhaven National 

Laboratory 

Document Number: ORAUT-TKBS-0048

Revision: 01 

Effective Date: 04/26/2010

Type of Document: TBD

Supersedes: Revision 00 

Subject Expert(s): Lori Arent, Samuel Chu, John A. Devanney, Leo Faust, Robert L. Morris, 

Eugene Potter, and Elyse M. Thomas

Site Expert(s): N/A

Approval: Signature on File Approval Date: 04/23/2010

Melton H. Chew, Document Owner

Concurrence: Signature on File Concurrence Date: 04/22/2010

John M. Byrne, Objective 1 Manager

Concurrence: Keith A. McCartney Signature on File for Concurrence Date: 04/23/2010

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 04/23/2010

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 04/26/2010

James W. Neton, Associate Director for Science

New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 2 of 115

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

08/30/2006 00 New Site Profile for the Brookhaven National Laboratory. First 

approved issue. Incorporates internal and NIOSH formal review 

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

Initiated by Melton H. Chew.

04/26/2010 01 Revision to incorporate SEC language, BNL comments, and clarifies 

that background radiation is not included in occupational exposure. 

References were consolidated at the end of the document. 

Incorporates formal internal and NIOSH review comments. 

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

determined by the Objective Manager. Initiated by Melton H. Chew.
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 3 of 115

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

7

1.0 Introduction ............................................................................................................................10

1.1 Purpose......................................................................................................................11

1.2 Scope.........................................................................................................................11

1.2.1 Site Description............................................................................................... 11

1.2.2 Occupational Medical Dose............................................................................. 12

1.2.3 Occupational Environmental Dose .................................................................. 12

1.2.4 Occupational Internal Dose............................................................................. 12

1.2.5 Occupational External Dosimetry .................................................................... 12

1.2.6 Attributions and Annotations ........................................................................... 12

2.0 Site Description......................................................................................................................13

2.1 Introduction.................................................................................................................13

2.2 Purpose......................................................................................................................13

2.3 Scope.........................................................................................................................13

2.4 Site Description and General Information ...................................................................13

2.4.1 Brookhaven Graphite Research Reactor, Building 701 ................................... 15

2.4.2 High Flux Beam Reactor, Building 750............................................................ 20

2.4.3 Medical Research Center, Building 490 .......................................................... 21

2.4.4 Brookhaven Medical Research Reactor, Building 491..................................... 21

2.4.5 Positron Emission Tomography, Building 906 ................................................. 22

2.4.6 Cosmotron, Building 902................................................................................. 22

2.4.7 Alternating Gradient Synchrotron, 900 Series Buildings .................................. 22

2.4.8 Brookhaven LINAC Isotope Producer, Building 931 ........................................ 27

2.4.9 Tandem Van de Graaff, Building 901 .............................................................. 27

2.4.10 National Synchrotron Light Source, Building 725 ............................................ 29

2.4.11 Relativistic Heavy Ion Collider Complex, Building 1000 Series........................ 29

2.4.12 Hot Laboratory, Building 801........................................................................... 31

2.4.13 High Intensity Radiation Development Laboratory, Building 830 ..................... 33

2.4.14 Chemistry Department, Building 555............................................................... 33

2.4.15 Physics Department, Building 510................................................................... 34

2.4.16 Other Miscellaneous Radiological Research Facilities .................................... 34

2.4.16.1 LMFR Support, Building 820............................................................. 34

2.4.16.2 Hot Machine Shop, Building 530....................................................... 34

2.4.16.3 Health Physics and Safety Instrumentation, Building 535 ................. 35

2.4.17 Waste Management Facility, Radwaste Reclamation Building, 

Building 865 .................................................................................................... 35

2.4.18 Radiological Waste Decontamination Facility, Building 650............................. 36

2.5 Isotopes of Concern ...................................................................................................36

2.6 Magnitude of Site Activity ...........................................................................................41

2.7 Major Incidents ...........................................................................................................46

2.8 Radiological Access Controls .....................................................................................47

3.0 Occupational Medical Dose.................................................................................................... 48

3.1 Purpose......................................................................................................................48

3.2 Scope.........................................................................................................................48

3.3 Examination Frequency..............................................................................................48

3.4 Equipment and Techniques ........................................................................................49
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3.5 Entrance Air Kerma in Air ...........................................................................................50

3.6 Organ Dose Calculations............................................................................................52

3.6.1 Organ Dose from Collimated Exposure........................................................... 52

3.6.2 Organ Dose from Poorly Collimated Exposure................................................ 52

3.6.3 Skin Dose ....................................................................................................... 53

3.7 Dose Reconstruction ..................................................................................................54

3.8 Uncertainty .................................................................................................................54

4.0 Occupational Environmental Dose ......................................................................................... 59

4.1 Introduction.................................................................................................................59

4.2 Purpose......................................................................................................................59

4.3 Scope.........................................................................................................................59

4.4 Ambient External Radiation ........................................................................................59

4.4.1 Use of Offsite Monitoring................................................................................. 59

4.4.2 Applicability for Dose Reconstruction.............................................................. 60

4.5 Inhalation of Onsite Airborne Radionuclides ...............................................................60

4.5.1 Stack Monitoring ............................................................................................. 60

4.5.2 Estimates of Potential Inhalation Intake .......................................................... 61

4.6 Ingestion.....................................................................................................................61

4.7 Uncertainty .................................................................................................................65

5.0 Occupational Internal Dose.................................................................................................... 66

5.1 Introduction.................................................................................................................66

5.2 Purpose......................................................................................................................66

5.3 Scope.........................................................................................................................66

5.4 Bioassay Records in the Individual Files.....................................................................66

5.4.1 Bioassay Record (Form 1720) ........................................................................ 66

5.4.2 Tritium Exposure Evaluation Form .................................................................. 67

5.4.3 Summary of Whole-Body Counting Results .................................................... 67

5.4.4 Individual Whole-Body Counting Results......................................................... 67

5.5 In Vitro Minimum Detectable Activities, Counting Methods, and Reporting 

Protocols ....................................................................................................................68

5.5.1 In Vitro Urine Analysis..................................................................................... 68

5.5.2 History of In Vitro Urine Analysis..................................................................... 68

5.5.3 Fecal Sample Analysis.................................................................................... 69

5.6 In Vivo Minimum Detectable Activities, Counting Methods, and Reporting 

Practices ....................................................................................................................70

5.6.1 Whole-Body Counting ..................................................................................... 70

5.6.2 Chest Counting ............................................................................................... 71

5.7 Uncertainty .................................................................................................................71

5.8 Detection Limits..........................................................................................................72

5.9 Excreta Sample Kit Codes..........................................................................................72

5.10 Solubility Type, Fraction Activity, and Particle Size by Facility .................................... 74

5.11 Facility-Specific Radionuclide Conversions.................................................................78

5.12 Workplace Monitoring Data ........................................................................................79

5.13 Source Term Data ......................................................................................................79

5.14 Radon.........................................................................................................................79

5.15 Guidance for Dose Reconstructors.............................................................................79

6.0 Occupational External Dosimetry ........................................................................................... 80

6.1 Introduction.................................................................................................................80

6.2 Purpose......................................................................................................................80
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6.3 Scope.........................................................................................................................80

6.4 BNL Dosimetry Program.............................................................................................80

6.4.1 Neutron Dosimetry .......................................................................................... 80

6.4.2 Special Dosimetry ........................................................................................... 80

6.4.3 Lost or Destroyed Dosimeters......................................................................... 81

6.5 Missed Dose...............................................................................................................81

6.5.1 Unmonitored Worker....................................................................................... 81

6.6 Radiation Energies and Percentages at Selected BNL Facilities ................................ 82

6.7 Neutron Radiations and Percentages .........................................................................83

6.8 Recorded Dose Practices ...........................................................................................85

6.9 Interpretation of Reported Doses................................................................................85

6.10 Adjustments to Recorded Dose ..................................................................................86

6.11 Bias and Uncertainty ..................................................................................................86

7.0 Attributions and Annotations .................................................................................................. 89

References ........................................................................................................................................91

Glossary .......................................................................................................................................... 105
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 6 of 115

LIST OF TABLES

TABLE TITLE

2-1 Site information of major nuclear research facilities at BNL....................................................

PAGE

17

2-2 Area information and parameters........................................................................................... 36

2-3 Magnitude of site activity........................................................................................................ 41

2-4 Major incidents.......................................................................................................................46

3-1 Default frequency of chest X-rays at BNL............................................................................... 49

3-2 Organ doses for PFG, PA, and LAT 14-in. × 17-in. chest radiography ................................... 54

3-3 Skin dose guidance for various chest projections and periods ............................................... 55

3-4 Skin dose from various chest projections ............................................................................... 57

4-1 External gamma radiation dose.............................................................................................. 60

4-2 Maximum annual environmental occupational radionuclide inhalation.................................... 62

5-1 Internal dose control program (in vitro)................................................................................... 69

5-2 Internal dose control program (in vivo) ................................................................................... 71

5-3 Detection limits.......................................................................................................................72

5-4 Excreta sample kit codes ....................................................................................................... 74

5-5 Solubility type, fraction activity, and particle size by facility..................................................... 75

5-6 Facility-specific radionuclide conversions............................................................................... 78

6-1 Estimated annual missed photon, beta, and neutron dose ..................................................... 81

6-2 Selection of beta and photon energies and percentages........................................................ 82

6-3 Beta and photon energies and percentages (support facilities) .............................................. 82

6-4 Facilities, neutron energies, percentages, and correction factors ........................................... 83

6-5 Recorded dose practices........................................................................................................ 85

6-6 Interpretation of reported data................................................................................................ 86

6-7 Adjustments to recorded dose................................................................................................ 86

6-8 Uncertainty factors .................................................................................................................87

6-9 Landauer gamma film, NTA film, and CR-39 specifications.................................................... 87

LIST OF FIGURES

FIGURE TITLE

2-1 Aerial view of Brookhaven National Laboratory ......................................................................

PAGE

14

2-2 Nuclear and radioactivity sites at BNL.................................................................................... 16

2-3 The AGS complex..................................................................................................................23

2-4 The Relativistic Heavy Ion Collider complex........................................................................... 30

2-5 RHIC ring intersection points.................................................................................................. 31

6-1 Neutron Spectrum in the Thermal Column with the Converter Plate (left) 

and in the Instrument Tunnel (right) ....................................................................................... 84

6-2 Neutron Energy Spectrum at the Cosmotron.......................................................................... 84
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 7 of 115

ACRONYMS AND ABBREVIATIONS

ADD average depth dose

AEC U.S. Atomic Energy Commission

AF absorption factor

AGS Alternating Gradient Synchrotron

AMAD activity median aerodynamic diameter

amu atomic mass unit

AP anterior-posterior

BF Backscatter factor

BGRR Brookhaven Graphite Research Reactor

BLAF Brookhaven Laboratory Animal Facility

BLIP Brookhaven LINAC Isotope Producer

BMRR Brookhaven Medical Research Reactor

BNL Brookhaven National Laboratory

Bq becquerel

cGy centigray

Ci curie

CLIF Chemistry LINAC Irradiation Facility (also CLIP in some documents)

cm centimeter

CNF Cold Neutron Facility

CR-39 Columbia Resin Number 39

d day

D&D decontamination and decommissioning

DAS Department of Applied Science

DAT Department of Applied Technology

DCF dose conversion factor

DTL Drift-Tube-LINACS

DOE U.S. Department of Energy

DOELAP Department of Energy Laboratory Accreditation Program

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EML Environmental Measurement Laboratory

ENSD entrance skin dose

ENSDNPB entrance skin dose near primary beam

ESE entrance skin exposure

EXSD exit skin dose

EXSDNPB exit skin dose near primary beam

F fast (solubility rate)

ft foot

g gram

GeV gigaelectron-volt

GIF Gamma Irradiation Facility

HEBT High Energy Beam Transport

HEPA high-efficiency particulate air (filter)

HFBR High Flux Beam Reactor
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 8 of 115

HIRDL High Intensity Radiation Development Laboratory

hr hour

HVL half-value layer

ICRP International Commission on Radiological Protection

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt (1,000 electron-volts)

kV kilovolt

kVp kilovolts-peak

L liter

LAT lateral 

lb pound

LEAF Low Energy Accelerator Facility

LINAC linear accelerator

LMFR Liquid Metal Fuel Reactor

M moderate (solubility rate)

m meter

mAs milliampere-second

mCi millicurie

MDA minimum detectable activity or amount

MDL minimum detectable level

MEL Metallurgical Evaluation Laboratory

MeV megaelectron-volt, 1 million electron-volts

MFP mixed fission product

min minute

mL milliliter

mm millimeter

mR milliroentgen

mrad millirad

MRC Medical Research Center

mrem millirem

MW megawatt

NaI(Tl) Sodium iodide doped with thallium

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

NIOSH National Institute for Occupational Safety and Health

NSLS National Synchrotron Light Source

NSRL NASA Space Radiation Laboratory

NTA nuclear track emulsion, type A

NVLAP National Voluntary Laboratory Accreditation Program

NYOO New York Operations Office 

ORAU Oak Ridge Associated Universities

OW open window

PA posterior-anterior

pCi picocurie

PET positron emission tomography
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PFG photofluorography

POC probability of causation

QF quality factor

R roentgen

R&D research and development

RARAF Radiological Research Accelerator Facility

RBE relative biological effectiveness

RF radio frequency

RFQ Radio-Frequency Quadrupole

RHIC Relativistic Heavy Ion Collider

RSD remote skin dose

RTF Radiation Therapy Facility

S slow (solubility rate)

s second

S&EP Safety and Environmental Protection (Division)

SEC Special Exposure Cohort 

ST Severn-Trent

TBD technical basis document

TPL Target Processing Laboratory

U.S.C. United States Code

VUV Vacuum Ultraviolet

WB whole-body

WBC whole-body counter

WMF Waste Management Facility

yr year

μA microampere

μCi microcurie

μg microgram

μm micrometer

§ section or sections
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 10 of 115

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

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2007a):

] guidelines established under 

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

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

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

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

conventional structures

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

reasons

 1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 11 of 115

1.1 PURPOSE

The purpose of this document is to provide a site profile that contains technical basis information for 

the evaluation of the total occupational dose for EEOICPA claimants who were employed at the 

Brookhaven National Laboratory (BNL).

1.2 SCOPE

This site profile consists of the following sections: 1.0, Introduction; 2.0, Site Description; 3.0, 

Occupational Medical Dose; 4.0, Occupational Environmental Dose; 5.0 Occupational Internal Dose; 

6.0, Occupational External Dosimetry; and 7.0, Attributions and Annotations.

A Special Exposure Cohort (SEC) class has been established for all employees who worked in any 

area of BNL from January 1, 1947, through December 31, 1979, for a number of workdays 

aggregating at least 250 workdays or in combination with workdays within the parameters established 

for one or more classes of employees in the SEC. This document also provides guidance for 

EEOICPA-covered employees who participated in BNL operations, specifically for non-SEC cancers 

and those presumptive cancer claims for workers who have fewer than 250 workdays under this 

employment or in combination with workdays within the parameters established for other classes of 

employees in the SEC. Based on the findings and recommendations of NIOSH and the Advisory 

Board on Radiation and Worker Health, the Secretary of Health and Human Services has concurred 

with the finding that NIOSH does not have access to sufficient personnel or area monitoring data, or 

sufficient source or source term information, associated with BNL operations to bound potential 

internal exposures for the period from January 1, 1947, through December 31, 1979 (other than 

tritium after December 31, 1964) (Sebelius 2009).

NIOSH has found that it does have access to sufficient personnel monitoring and workplace 

monitoring data to bound potential external exposures and occupational medical dose for workers at 

BNL during the period from January 1, 1947, through December 31, 2007.

NIOSH found that while it is not possible to completely reconstruct radiation doses for employees who 

worked at BNL from January 1, 1947, through December 31, 1979, NIOSH intends to use any reliable 

internal and external monitoring data that may be available for an individual claim (and that can be 

interpreted using existing NIOSH dose reconstruction processes or procedures) to support a partial 

dose reconstruction for nonpresumptive cancers and/or cases that have less than 250 work days of 

employment.

1.2.1

Section 2.0 of this site profile describes the features and the history of BNL, including site areas or 

buildings, site processes, periods of operation, radionuclides of concern, and other information 

pertinent to dose reconstruction. 

Site Description

This section provides a description and general information for the facilities that operated or are 

operating at BNL. These include the Brookhaven Graphite Research Reactor (BGRR) and its 

replacement, the High Flux Beam Reactor (HFBR); the operations of the Medical Research Center

(MRC), the positron emission tomography facility and the medical research reactor; the operation of 

the accelerators and the production of radiopharmaceuticals. The section also addresses some of the 

high-energy physics facilities such as the relativistic heavy ion collider complex where ion beams 

traveling in opposite directions in a circular orbit collide.
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 12 of 115

1.2.2

Section 3.0 of this site profile provides information regarding the historical BNL occupational X-ray

program. Beginning in 1947 and continuing through at least 1964, workers, including visiting 

scientists employed for more than three months, were given an entry and annual physical examination 

including a posterior-anterior (PA) chest X-ray. Later, the annual frequency may have actually been 

closer to about once every 2.5 years. The section presents organ doses for both PA and lateral (LAT)

chest radiographs assuming both properly collimated and poorly collimated exposures. Skin doses 

inside and outside the primary beam are determined for PA and LAT chest projections.

Occupational Medical Dose

1.2.3

Section 4.0 discusses the radiation doses received by workers at BNL but outside the facilities. From 

1967 to 1984, ambient gamma radiation dose rates are available from four perimeter stations 

surrounding the site at northwest, southwest, southeast, and northeast coordinates. Beginning in 

1985, the onsite monitoring program was expanded to approximately 20 onsite stations. The ambient 

external gamma radiation outside the radiological controlled areas is the result of background 

radiation or gamma emissions from stack effluent such as 41Ar or skyshine due to air scatter from an 

otherwise well-shielded radiation source. The radionuclides emitted from the BNL stacks are 

identified and their annual inhalation (becquerels/year) listed. Neither potable groundwater nor soil 

ingestion has been found to be a pathway of exposure. No ingestion dose is indicated.

Occupational Environmental Dose

1.2.4

Section 5.0 discusses the internal dosimetry systems and practices employed at BNL. The section 

provides supporting technical data to evaluate internal occupational doses that can reasonably be 

associated with worker radiation exposures covered by EEOICPA. It lists the types of routine 

monitoring that constitute the in vitro and the in vivo dose control programs as well as the minimum 

detectable activities, counting methods, and reporting practices. The section addresses the 

uncertainty for BNL exposure and dose records.

Occupational Internal Dose

1.2.5

Section 6.0 discusses external dosimetry systems and practices at BNL. It presents historical and 

current practices in relation to the evaluation of external exposure data for monitored and unmonitored 

workers. It contains supporting documentation to assist in the evaluation of occupational external 

exposures from processes that occurred at BNL. In addition, it addresses the evaluation of worker 

exposure, missed dose, and the bias and uncertainty associated with the monitoring of external dose. 

The missed doses were basically those resulting from dosimeter minimum detectable levels (MDLs) 

and exchange periods, as well as the inability to determine true doses from high-energy particles 

generated by the accelerators. Neutron doses were measured with nuclear track emulsion, type A 

(NTA) film until 1995, after which Columbia Resin No. 39 (CR-39) has been used.

Occupational External Dosimetry

1.2.6

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.

Attributions and Annotations
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2.0

2.1 INTRODUCTION

SITE DESCRIPTION

BNL was founded in 1947, and it has been in operation since then at Upton, Long Island, New York. 

The site was formerly Camp Upton and it was used by the Army during World Wars I and II.

BNL’s early research focused on advanced physics, but expanded into its current suite of research in 

the fields of medicine, biology, chemistry, physics, materials science, nuclear engineering, and 

environmental research. BNL was established to provide facilities for scientific research and was 

organized into departments that provide research nuclear reactors, particle accelerators, and 

engineering facilities, and that support the Biology, Chemistry, Physics, Medicine, Applied Science, 

Accelerator, and Applied Mathematics Departments.

2.2 PURPOSE

The purpose of this section is to describe the BNL facilities, the time and period of the process 

operations, and major site incidents that resulted in a significant internal or external exposure to one 

or more persons.

2.3 SCOPE

This section describes the features and history of BNL, including the site areas or buildings, site 

processes, periods of operation, radionuclides of concern, and other information pertinent to dose 

reconstruction.

2.4 SITE DESCRIPTION AND GENERAL INFORMATION

Figure 2-1 shows an aerial view of BNL. 

Research Reactors

The Laboratory’s scientific history began in 1950 with the operation of the BGRR, a research reactor 

used for peaceful scientific exploration in the fields of medicine, biology, chemistry, physics, and 

nuclear engineering. The BGRR operated until 1968. In 1965, its capacity was replaced and 

surpassed by the HFBR, which provided neutrons to researchers of all disciplines, from solid state 

physics to art history. The HFBR ceased operations in December 1996.

Medical Research Center 

Medical research at BNL began in 1950 with the opening of one of the first hospitals devoted to 

nuclear medicine. It was followed by the MRC in 1958 and the Brookhaven Medical Research 

Reactor (BMRR) in 1959. The BMRR is the first nuclear reactor built exclusively for medical and 

biological research and came on line on March 15, 1959, and operated until October 2000. 

The Radiation Therapy Facility (RTF) was operated jointly by the BNL Medical Department and the 

State University of New York at Stony Brook. The RTF was a high-energy, dual X-ray mode linear 

accelerator (LINAC) for radiation therapy of cancer patients. The RTF has been dismantled and is not 

operational.

Chemists and physicians teamed to view the inner workings of the brain in 1977 with the advent of 

Positron Emission Tomography (PET) cameras. Two more imaging techniques were added to the 

PET research efforts to form the Center for Imaging and Neuroscience in 1996. Except for the 

BMRR, all of these facilities are currently operating. 
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Figure 2-1. Aerial view of Brookhaven National Laboratory (BNL 2009a).

Particle Accelerators

High-energy particle physics research began in 1952 with the Cosmotron, the first particle physics 

accelerator to achieve billion-electro-volt energies. The Cosmotron operated from 1953 to 1966. In 

1960, the Alternating Gradient Synchrotron (AGS), a large accelerator, was built to surpass the 

Cosmotron’s capabilities. The AGS is capable of accelerating protons to energies up to 30 GeV and 

heavy ion beams to 15 GeV/amu. The AGS achieved full energy in 1960 and is still in use. 

Between 1967 and 1970, the Tandem Van De Graaff, 60-inch Cyclotron, and Vertical Accelerator

were used for medium-energy physics investigations and for special isotope production. The heavy 

ions from the Tandem Van De Graaff can be injected into the AGS for physics experiments. The 

Tandem Van De Graaff began operating in 1970 and continues operating to the present.

The Heavy Ion Transfer tunnel connects the coupled Tandem Van de Graaff and the AGS. The 

interconnection of these two facilities permits intermediate-mass ions to be injected into the AGS 

where they can be accelerated to an energy of 15 GeV/amu. These ions then are extracted and sent 

to the AGS experimental area for physics research. The AGS Booster is a circular accelerator with a 

circumference of 200 m that receives either a proton beam from LINAC or heavy ions from the 

Tandem Van De Graaff. The Booster accelerates proton particles and heavy ions before injecting 

them into the AGS ring. The Booster receives protons and heavy ions from the LINAC and Tandem 

Van De Graaff facilities to increase their intensity for delivery to the AGS.

The Brookhaven LINAC Isotope Producer (BLIP) became operational in 1973. Protons from the 

LINAC are sent through an underground beam tunnel to the BLIP facility where they strike target 
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metals. These metals, which become activated by the proton beam, are then processed at the Target 

Processing Laboratory for use in radiopharmaceutical development and production. The targets are 

cooled by a continuously recirculating water system. The BLIP facility underwent significant upgrades 

in 1994 in support of the Brookhaven Isotope Research Center program. The 200-MeV Proton Linear 

Accelerator serves as a proton injector for the AGS and supplies a continuous beam of protons for 

radionuclide production by spallation reactions in the BLIP.

In 1982, the National Synchrotron Light Source (NSLS) began operation. The NSLS guides charged 

particles in an orbit. As the electrons spin inside a hollow donut-shaped tube called an electron 

storage ring, they emit light called synchrotron light. This light, which can be detected by specialized 

instruments, is used to study the properties of matter. The NSLS utilizes a LINAC and booster 

synchrotron as an injection system for two electron storage rings that operate at energies of 750 MeV 

vacuum ultraviolet (VUV), and 2.5 GeV (X-ray). The synchrotron radiation produced by their stored 

electrons is used for VUV spectroscopy and for X-ray diffraction studies.

Brookhaven’s newest accelerator facility is the Relativistic Heavy Ion Collider (RHIC), which was 

completed in 1999. The RHIC is designed to recreate a state of matter that scientists believe existed 

moments after the universe was formed. 

Department of Applied Science 

The Target Processing Laboratory (also call Hot Laboratory in 1993) officially opened on January 15, 

1951. The original purpose of the central facility was to provide appropriately shielded areas for 

research with large amounts of radioactive material. The “hot” area of the Hot Laboratory included 

five hot cells, three chemical-processing hot cells, and three high-level hot cells for handling and 

processing of radioactivity in gaseous, liquid, or solid form.

In 1971 and 1972, the High Intensity Radiation Development Laboratory (HIRDL), which contained 

million-curie-range (in the early 1970s) 60Co and 137Cs sources, was used for source development and 

experimental process irradiations. A 60Co pool in the HIRDL facility operated at lower activity levels 

into the 1990s.

Waste Management

The new Waste Management Facility (WMF) replaces the original Hazardous Waste Management 

Facility in its entirety and consolidates several waste management operations into functional 

buildings. The WMF also provides for significant expansion in a dedicated site suitable for handling 

and storing hazardous and radioactive wastes generated at BNL. 

Figure 2-2 shows the location of the major sites where nuclear research activities were carried out at 

BNL.

Table 2-1 lists the major nuclear research facilities and their associated summary information.

The following subsections comprise further descriptions of the research facilities of significance on the 

BNL site and provide background information relevant to the dose reconstruction efforts.

2.4.1

The operational BGRR and its support systems were housed in Buildings 701, 704, 708, and 709.

Brookhaven Graphite Research Reactor, Building 701

Building 701

The BGRR was a graphite-moderated and -reflected, thermal-neutron, air-cooled research facility. 

The original fuel loading was natural uranium, which was replaced in later years by highly enriched 

fuel elements. The 25-ft, 700-ton graphite cube was built in two halves separated by a narrow vertical 
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 16 of 115

Hot 

Laboratory

RHIC

Linac

Booster Accelerator

Alternating

Gradient

Synchrotron

Tandem

Van de Graaff

Medical 

Research 

Center 

(BMRR)

BGRR

HFBR

Chemistry 

Department

BLIP

Cosmotron

PET

NSLS

HIRDL

Physics 

Department

Figure 2-2. Nuclear and radioactivity sites at BNL [1].
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 17 of 115

Table 2-1. Site information of major nuclear research facilities at BNL.

Section Name Description Period 

Building 

No. Note 

2.4.1 BGRR U-fueled, graphitemoderated and -reflected 

reactor

28 MW

1950–1968 701

2.4.1 BGRR fuel canal BGRR fuel storage 1950–1972 709

2.4.1 BGRR exhaust 

fanhouse

Highly contaminated ducting 

and fans

1950–1972 704

BGRR exhaust 

filter house

Contaminated HEPA filter 

elements

1950–

unknown

708 Filter elements were 

still in place in 1997.

2.4.2 HFBR 30–60 MW thermal heavywater-moderated research 

reactor

1965–1999 750 Building ceased

operation due to 

leak in spent fuel 

pool in 1997.

2.4.2 HFBR Stack 100-m stack and houses 

iodine filters

1965–1999 705

2.4.2 CNF Produce extremely lowenergy neutron by moderate 

neutron with liquid hydrogen

1980–1996 751

2.4.3 MRC Medical Department 1959–present 490 Dispersible 

radioisotopes 

PuBe sources 

(removed)

2.4.4 BMRR Enriched U-fueled, lightwater-moderated and 

-cooled reactor

1959–2000 491

2.4.5 PET Imaging 

Laboratory

PET scanners for nuclear 

medicine and neurosciences 

researches

1977–present 906 Use of positron 

emitters (carbon-11,

fluorine-18, nitrogen13, and oxygen-18)

2.4.6 Cosmotron Proton accelerator 1949–1968 902

2.4.7 AGS A large accelerator capable 

of accelerating protons to 

energies up to 30 GeV and 

heavy ion beams to 15 

GeV/amu

1960–present 913 AGS tunnel, shortlived activation 

products in tunnel.

2.4.7 200-MeV LINAC Preaccelerator and LINAC

providing proton source

1967–present 930

2.4.7 LINAC cooling 

support service

Water-cooling services for 

the LINAC complex

1967–present 930 MER Contains activated 

water

2.4.7 AGS Booster 

Tunnel

Accumulator booster ring 1960–present 942 Short-lived activation 

products in tunnel

2.4.7 AGS Experiment 

Area

Experimental target area for 

the AGS ring and later 

contains a proton beam 

switchyard system

1960–present 912 Activated 

components

2.4.7 AGS ventilation 

system fanhouses

Ventilation exhaust fans 1960–present 913A-E High radiation area 

due to proximity to 

the tunnel. 

Activated water

2.4.7 AGS LINAC Contains 50-MeV AGS 

LINAC

1960s 914
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Section Name Description Period 

Building 

No. Note 

2.4.7 E10 Power 

Supply Building

Houses an assortment of 

power supplies

1971–present 920 Radiation hazard 

from activated 

cooling water and oil 

in transformer.

2.4.7 Siemens motor 

generator house

Electrical equipment and 

lubricating equipment for the 

motor generator

1960–present 928, 929 Radiation hazard 

from activated 

cooling water

2.4.7 AGS Department Office, field testing shops, 

machine shops, high bays

1960–present 911 Stored and repair 

activated 

components.

2.4.7 g-2/Bubble 

Chamber

Contains the 30-in. and 80-

in. Bubble Chamber

1962–1978 919 A release of tritium 

occurred in 1973.

2.4.7 Proton Target Proton target that produces 

secondaries for the g-2 ring

1962–present 949 Highly activated 

target

2.4.7 7-ft Bubble 

Chamber 

High-energy physics 

experimental facility

1972–1979 960 Activated 

components

2.4.7 AGS Warehouses Warehouses 1960–present 975, 918,

196, 209,

424, 178

Stored activated 

components

2.4.8 BLIP Production of radioisotopes 

via irradiation in the LINAC 

beam (200-Mev protons)

1961–present 931B Bldgs. 931A and C 

have no nuclear 

hazard.

2.4.9 Tandem Van de 

Graaff 

Van de Graaff accelerators

60-in. proton Cyclotron 

producing 10-MeV protons, 

20- MeV deuterons, and 40-

MeV alpha particles for 

medical radionuclide 

research

1970–present 901 Dynamitron (positron 

accelerator) 

operations ceased in 

1989.

2.4.10 NSLS Produce intense sources of 

X-rays, ultraviolet and 

infrared radiation

1981–present 725

2.4.10 NSLS 

Development 

Laboratory

Houses a linear accelerator. 1981–present 729

2.4.11 RHIC Complex High-energy particle 

accelerator and collider

1999–present 1002 Has prior radioactive 

works.

BRAHMS experimental hall 1981–present 1004A Uranium plate 

machining

RHIC service building 1981–present 1004 Has prior radioactive 

Open area (future works.

experimental area)

1996–present 1004B

RHIC support building 1994–present 1005

Beam tunnel 1981–present 1005S

Offices and tech shops 1981–present 1006

STAR Detector 1985–present 1006A

Service building 1981–present 1007

Beam tunnel 1981–present 1008

PHENIX experiment 1988–present 1008A

Experimental support 

building

1985–present 1010

PHOBOS experiment 1994–present 1012

Future experimental hall 1994–present
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Section Name Description Period 

Building 

No. Note 

2.4.12 Hot Laboratory Target Processing 

Laboratory

BLIP processing group

1951–present 801 There was an 

explosion of 

UF6/BrF3 on 

5/15/1957. Few 

people injured.

2.4.13 HIRDL Radioactive sources 

fabrication, irradiation cells 

and 1970-1990s, the GIF 

1951–1990s

GIF 1970–

present

830 The GIF contained 

approximately 

35,000 Ci of Co-60 

(removed) and 

megacurie Co-60 

and Cs-173 sources 

(removed)

2.4.14 Chemistry 

Department

Studies from chemical 

kinetics, to hot-atom effects, 

to nuclear chemical studies 

of nuclear decay schemes 

and nuclear reaction crosssections, to solar neutrino 

studies and radiocarbon 

dating.

1966–present 555

2.4.15 Physics 

Department

Laboratories, office space, 

machine shop, and 

basement area

1962–present 510 Radioactive material 

storage, counting 

labs 

2.4.16.1 LMFR Support 

Facility

Research and development 

work for the LMFR

1957–1975 820

2.4.16.2 Hot Machine 

Shop

Hot Machine Shop 1947–1975 530

2.4.16.3 Instrument 

Division

Research and development 

of X-ray and neutron 

detectors and corresponding 

electronics, which are used 

at the NSLS, the HFBR, the 

AGS, and other worldwide 

scientific research facilities.

1964–present 535

2.4.17 Radwaste 

Reclamation 

Building

Located in the WMF for 

processing of radwaste

1997–present 865

2.4.17 Tritium 

Evaporator 

Facility

Reduce tritium release by 

evaporation.

(Closed) 802B

2.4.17 Waste 

Management 

Incinerator

Low-level waste incinerator 1970s–early 

1990s

444 Animal carcasses 

with no radiological 

content were 

incinerated.

2.4.18 Radiological 

Decontamination 

Facility

Decontamination of 

radiological items and Hot 

Laundry operated until 1989

1959–1996 650

gap running east-west. Filtered cooling air was drawn into this gap and flowed through north-south 

channels removing heat from the fuel elements and graphite, and then flowed out of the reactor 

through underground concrete ducts, passing through filters, coolers, and primary exhaust fans that 

discharged into a 100-m high stack (BNL 1997a, p. 3).

The BGRR achieved criticality on August 22, 1950, and operated at power levels up to 28 MW with 

natural uranium fuel. In April 1958, the reactor was reloaded with highly enriched uranium fuel 
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 20 of 115

elements and operated at power levels of up to 20 MW. On June 10, 1968, the use of the reactor for 

experimental purposes ended and initial decommissioning operations began. The last fuel was 

shipped to the Savannah River Site in June 1972 and the canal was pumped dry (BNL 1997a, p. 4). 

Building 709

Building 709 is the fuel transfer and storage canal of the BGRR. During the 1950s, before the 

replacement of the natural uranium fuel slugs, there was a substantial problem with corrosion of fuel 

elements in the canal. This fuel was subject to deterioration and oxidation while in storage, making 

the slugs “dusty” when they were eventually prepared for shipment. The new type of fuel elements, 

which replaced the natural uranium elements in 1958, eliminated this problem. The new elements 

were aluminum-clad enriched uranium.

During the early years of operation of the BGRR, fuel failures occurred that resulted in radioactive 

materials being released to the air stream that cooled the reactor. There were 28 reported ruptures of 

BGRR fuel during the period from 1952 to 1957. These all occurred with the natural uranium fuel. 

There was one rupture of a uranium oxide (U3O8) sample that was being irradiated for the radioiodine

production program. Aside from 41Ar, 131I was the most important radionuclide that would contribute to 

a potential dose that was discharged to the atmosphere from the BGRR. Bromine-82 and 133I were 

released in somewhat larger concentrations (Meinhold and Meinhold 2001).

More than 2,414 fuel elements, generated in a 12-year period, were shipped from the canal. The 

contaminated water, filter media, and backflushing from the ion exchange columns were pumped to 

the storage tanks at the Waste Concentration Facility, Building 811.

By 1972, the canal was finally drained, cleaned with detergents and water, and covered with concrete 

slabs for shielding (BNL 1997b, p. 8).

Building 704

The BGRR was shut down in 1968 and partially decommissioned in 1972. The exhaust ducts from 

the reactor and to the stack have been sealed from the fans. The fans remain in their cells and are 

contaminated. The intake duct, exhaust duct, and fans are grossly contaminated from fuel failures 

that occurred with the natural uranium fuel (BNL 1997b, p. 6). 

Building 708

Radioactive hot air was filtered and discharged from the reactor through underground concrete ducts 

between the reactor, Building 701, the instrument house, Building 708, and the fanhouse, Building 

704. The north and south exit air ducts were contaminated as a result of fuel ruptures with the original 

natural uranium fuel. Due to decay over some 30 years since use of uranium metal fuel, the major 

remaining fission products should be 137Cs and 90Sr. On August 18, 1988, a radiological survey was 

conducted on each exhaust duct upstream of the filters. Results indicated 20,000 to 30,000 dpm in 

the ducting. This indicates 40 to 70 times higher than the unrestricted limit of 500 dpm. Filter 

elements and exhaust coolers have been removed (BNL 1997b, p. 6).

2.4.2

The BGRR capacity was replaced and surpassed in 1965 by the HFBR (BNL 2001a). The HFBR is a 

30-MW thermal, heavy-water-moderated nuclear research reactor. The reactor is designed to provide

intense beams of neutron radiation to be used in the study of a variety of neutron scattering research 

projects and to allow for the radiating of materials close to the reactor core by insertion through 

special tubes near the reactor core. The HFBR is fueled with fully enriched uranium and aluminum 

alloy, moderated and cooled by heavy water. Initial criticality was achieved on October 31, 1965. The 

HFBR was originally designed to operate at 40 MW. In 1979, a newer fuel design was employed that 

High Flux Beam Reactor, Building 750
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 21 of 115

provided increased uranium loading in anticipation of an overall facility upgrade. In 1982, several 

major modifications were completed that allowed operation at powers up to 60 MW. 

In 1990, as a result of questions raised about the ability of natural circulation to cool the reactor during 

certain emergencies, reactor operations were limited to 30 MW. Heavy water flowing in the core is 

exposed to a dense neutron field that activates the deuterium atoms in the water to produce tritium. 

The primary mechanism by which tritium is transferred from the interior coolant system to the 

atmosphere is depressurization of the reactor vessel and evaporative losses during maintenance and 

refueling operations. During a scheduled maintenance shutdown in 1997, a leak in HFBR’s spent fuel 

storage pool was discovered. In November 1999, the Secretary of Energy made a decision to 

permanently close the HFBR. 

The HFBR Complex includes the following buildings:

Building 750

Building 750 is a domed cylindrical structure that contains the reactor and almost all associated 

process systems. The reactor, its beam lines, and laboratories are on the second main level known 

as the Experimental Level. The topmost level, the Operations Level, contains the Control Room and 

the Instrument Maintenance Shops.

Building 751

In 1980, the Cold Neutron Facility (CNF) was installed at beam line H9. This facility provided 

extremely-low-energy neutrons by moderating thermal neutrons with liquid hydrogen. Cooling of the 

liquid hydrogen was provided by liquid helium cooled by a large compressor in Building 751. 

Building 705

Building 705 is a 100-m stack that provides the path for ventilation exhaust from the HFBR. A 30-in. 

underground duct exhausts air from Building 750 ventilation blowers to particulate and iodine filters 

and then up through the stack. Stack effluent downstream of the filters is monitored from the Stack 

Monitoring Facility in Building 715 (BNL 1997c, pp. 6, 7). 

2.4.3

The Medical Department (Building 490) was opened in 1956 and conducted research in the building. 

Active clinical research involving human volunteers is conducted in the Clinical Research Center and 

its satellite facilities. Animal research is conducted in the Brookhaven Laboratory Animal Facility 

(BLAF). Benchtop laboratory research is conducted in support of the various programs. Studies 

involving dispersible radioisotopes (used primarily as tracers) are carried out in the building.

Medical Research Center, Building 490 

There are several other facilities in the building, among them the Whole Body Neutron Irradiation 

Facility, the Prompt Gamma Neutron Facility, the Partial Body Neutron Activation Facility, the Inelastic 

Neutron Scattering Facility, the Whole Body Counter, and two SPECT cameras. Some of these 

facilities contain radiation sources (e.g., PuBe sources (removed) for the Whole Body Neutron 

Irradiation Facility were stored in a vault below the facility and were raised into the facility when in 

use). In addition, there are many fume hoods and other apparatus typically found in research 

laboratories (BNL 1997d, pp. 4, 5). 

2.4.4

The BMRR, Building 491, is the first nuclear reactor built exclusively for medical and biological 

research and came on line on March 15, 1959, and operated until October 2000. It is an integral part 

of the MRC and is in a 60-ft-diameter steel and concrete structure adjacent to the MRC (Building 490). 

It is fueled with enriched uranium, moderated and cooled by light water, and operated intermittently at

Brookhaven Medical Research Reactor, Building 491
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 22 of 115

power levels up to 3 MW (thermal). The reactor is capable of operating for short periods at powers up 

to 5 MW. Air from the interior of the containment building is used to cool the neutron reflector 

surrounding the core of the BMRR vessel. When air is drawn through the reflector, it is exposed to a 

neutron field that causes the natural argon gas in the air to become radioactive (as 41Ar). Due to a 

reduction of research funding, the BMRR conducted its last run on December 2000, and transition and 

stabilization activities began in 2001. 

The research function of the reactor was to provide radiation beams of known, controllable character 

and strength at the designed sample and treatment locations. In 1997, its primary use was to provide 

neutron beams for Boron Neutron Capture Therapy research and patient treatment. Testing of 

neutron capture compounds and techniques was ongoing (BNL 1997e, p. 4). 

2.4.5

Building 906 was constructed in 1981 and houses two state-of-the-art PET scanners for use in basic 

research in the fields of neurology, nuclear medicine, and neurosciences.

Positron Emission Tomography, Building 906

Radiopharmaceuticals labeled with short-lived positron emitters (11C, 18F, 13N, and 15O) are used for 

physiological and biochemical research.

In 1987, the research group acquired a new PET scanner and another in 1997. The name changes 

for Building 906 have been evolutionary in that today it is called the Pet Imaging Laboratory (BNL 

1997f, p. 5).

2.4.6

The 902 Complex was originally built for the Cosmotron proton accelerator and experimental facilities. 

Some of the areas date back to Camp Upton. The Cosmotron was a very-low-intensity accelerator.

Cosmotron, Building 902

After the Cosmotron ceased operation, the Medical Department continued the operation of the Van de 

Graaff accelerator for a radiobiological program. The balance of the high bay area of Building 902 

was used for research and development of superconducting magnets. Eventually, the Medical 

Department moved its equipment out and the high bay was devoted entirely to the development and 

fabrication of a superconducting magnet.

The operational chronology of the Building 902 High Bay is:

1949–1968 Cosmotron

1968–1980 Radiation effects facility in the south portion of the high bay

1968–present Superconducting magnet development and fabrication and other equipment 

assembly (BNL 1997g, p. 2)

2.4.7

The AGS achieved full energy in 1960 and was capable of accelerating protons to energies up to 30 

GeV and heavy ion beams to 15 GeV/amu. When such accelerated particles collide and interact with 

nuclei, part of their energy is transformed into new particles, which fly off from the target nuclei. 

These particles are then detected or their paths are made visible by devices such as scintillation 

counters, Cerenkov counters, bubble chambers, spark chambers, and photographs. The AGS has 

been continuously upgraded and it now provides the highest intensity of protons per pulse and the 

highest intensity of polarized protons per pulse than any other synchrotron in the world. The AGS 

also accelerates heavy ion beams for experimental use as well as for the RHIC. 

Alternating Gradient Synchrotron, 900 Series Buildings 
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The AGS consists of three different accelerating systems: the 200-MeV LINAC, which is the source of 

the protons housed in Building 930, the 200-m-circumference rapid cycling Booster synchrotron in 

Buildings 914 and 942, and the 800-m AGS main ring designated as Building 913. The Tandem Van 

de Graaff accelerator, which is a separate facility, provides heavy ions for the Booster ring through a 

beam transfer line. The particles that are accelerated by the AGS facility are extracted into and 

stopped in various experimental areas around the ring, or are sent through a transfer line to the RHIC. 

Building 912 was the original experimental target area. The asphalt pad inside the AGS ring adjacent 

to Building 912 was an experimental area in the 1960s with two secondary beam lines. Since that 

time, activated components have been stored and repaired in the area. Figure 2-3 shows the 

buildings associated with the AGS Complex. 

Figure 2-3. The AGS complex (BNL 1997h). 
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 24 of 115

The AGS Linear Accelerator, Building 930 and Associated Facilities

The LINAC was designed and built in the late 1960s as part of a major upgrade to the AGS facility. 

The LINAC tunnel is 140 m long and consists of nine large accelerators. From the end of the LINAC,

a 17-m-long spur tunnel injects the beam into the Booster accelerator. Another high-energy beam 

transport (HEBT) line extends 95 m to connect with the AGS. Another spur tunnel provides beam to 

the BLIP facility, the Chemistry LINAC Isotopes Facility (CLIF), the Radiation Effects Facility, the 

Neutral Beam Test Facility, and a beam stop. Only the BLIP facility is presently used, and it has 

recently undergone a significant upgrade to increase its intensity. The LINAC consistently produces a 

beam of 25 to 35 mA with a 0.5-ms pulse length or 1.0 × 1014 protons/pulse. Pulses are repeated at 

the rate of 7.5 per second, but at most only 1 in 7 pulses are transported to the Booster with the 

remainder transported to the BLIP facility. 

Relevant buildings directly attached with building 930 are:

• Building 930 MER - provides water-cooling services for the LINAC complex and contains 

activated water. Associated with it is a large water-cooling tower and a compressor area. 

• Building 930 Annex - was added in the 1970s and contains shops and storage areas. Building 

930A was added on in the late 1980s as a service building for the Booster project. It contains 

power supplies and computer control equipment for the Booster (BNL 1997h).

The AGS Accelerator Rings, Buildings 913 and 914, and Associated Facilities

The AGS tunnel, Building 913, has a circumference of 800 m and is buried under a 7-m-thick berm of 

earth and concrete. There are 240 large alternating gradient dipole electromagnets, more than 100

supporting focus and beam tuning magnets as well as radiofrequency (RF) accelerating components, 

vacuum systems, water-cooling systems, etc., in the tunnel. Supporting the tunnel operation are the 

ventilation system fanhouses and multiple power supply houses. Some of these houses are 

designated 913A through R, while others have a separate plant engineering building number. The 

AGS tunnel (Building 913) is interrupted at the Experimental Hall, Building 912, for roughly 65 m. 

Here the AGS is under heavy concrete shielding blocks. 

The Booster tunnel (Building 942) is directly adjacent to the AGS tunnel and the LINAC. It is 

connected to the LINAC and the Tandem Van de Graaff through the HEBT transfer tunnel. The 

original connecting tunnel is now used to transfer a beam from the Booster to the AGS. The Booster 

tunnel is 200 m in circumference and is buried under a 5-m thick berm. 

The new section of the Booster tunnel was built in 1988. Some radioactive gases were created by 

activation of the ventilating air in the tunnel and experimental areas of the AGS. But, because they 

are short-lived (11C, half-life 20.5 minutes; 13N, half-life 10.0 minutes; 15O, 2.1 minutes), they did not 

create a measurable radiation level beyond the immediately adjacent onsite area. All of the cooling 

water systems in the AGS and Booster tunnels are activated. This includes systems that use the 

AGS ring as a utility tunnel for piping that goes to the support buildings. In addition, there are 

activated components in the AGS and Booster rings, and activated spare parts are stored in the 

facility (BNL 1997h).

Other Related Buildings

Building 911. Building 911 is the main office for the AGS Department. It was one of the first 

structures built for the accelerator complex and has direct passageways to both the AGS tunnel 

(Building 913) and the AGS experimental area (Building 912). Building 911 also contains the main 

control room, the support building for the main magnet cooling system, the support building for the 

Westinghouse main magnet power supply (now the backup power supply for the ring magnets), a high 

bay assembly area with magnetic field testing shops, and other light electronics, vacuum, and 

machine shops. There are activated components from the AGS ring that have been stored and 
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repaired in the high bay area and shop areas of Building 911 in posted radiation work areas. Because 

it also contains a system to cool water from the AGS ring, it does contain activated water (BNL 

1997h). 

Building 912

Within this large area, a majority of the ~400 AGS experiments have been located. The AGS 

experimental floor also contains a section of the AGS ring. Surrounding the beam lines are concrete 

and steel shielding needed to maintain radiation levels to within the required limits outside the lines. 

Shielding around the proton target stations that produce secondary beams is much thicker due to the 

high levels of radiation generated in the targets. This shielding can have significant activation 

contained in the body of the material. Cooling water at these target stations has higher levels of 

activation than the magnet cooling water and is handled by closed cooling systems. Radiation safety 

is provided by security systems that prevent access to these target stations and primary beam lines 

when protons are being extracted to these areas. There are activated components in the beam lines,

including highly activated targets. Steel cutting, including activated shielding steel, has been 

performed inside and outside Building 912. Depleted uranium was used in experimental areas in the 

past, but its use and disposition were tightly controlled. Small radioactive sources (nanocuries) are 

used. They are inventoried and stored in locked shielding boxes (BNL 1997h). 

. Building 912 is the designation of the approximately 5 indoor acres of AGS 

experimental floor. It is not one building, but five connected structures constructed over a 30-year 

period. The initial portion of Building 912 was started in 1958 to enclose both a portion of the AGS 

ring and the first experiments. 

Building 913A-E. AGS ventilation system fanhouses were an original part of the AGS ring and 

provided air conditioning and heating for the tunnel enclosure. Because of their proximity to the 

tunnel, they are High Radiation Areas when there is beam in the AGS ring and must be kept secured 

during operations. The cooling water for the air-conditioning system is activated and there is a 

potential for release in the fanhouse. Air handled in the fanhouses is from the AGS tunnel air and can 

also be activated (BNL 1997h).

Building 914. Building 914 was the location for the original 50-Mev AGS LINAC. This facility 

contained power supplies, water-cooling systems, pumps and compressors, and a motor generator 

set. During the 1970s and 1980s, after the LINAC was decommissioned, this building housed a 

mechanical maintenance group for the AGS division. Activated components were repaired and stored 

in this facility and new components were assembled and cleaned there (BNL 1997h). 

Building 919 (the g-2/Bubble Chamber and associated facilities). Building 919 was built in 1962 for an 

AGS experimental area and is still in use. The high bay area of Building 919 housed the 80–in. 

bubble chamber that was in operation until 1978. The bubble chamber had both a cryogenic 

component and a large mechanical component driven by a high-pressure oil hydraulic system. When 

the bubble chamber reached the end of its run, the high bay area and control room were used as 

shops and assembly areas. The cryogenic shop continued to be used to support liquid H2 target in 

Building 912 and the other buildings became storage areas. This continued to the late 1980s, when 

the area was modified for the g-2 experiment and the water-cooling systems were used to support 

Booster operations. A release of tritium from a gas chronograph detector occurred in 1973 in the 

technician machine shop. The detector held approximately 250 mCi of tritium (BNL 1997h). 

Building 920. Building 920 is the EI0 Power Supply building, and is above the AGS ring. It houses an 

assortment of power supplies used under different beam configurations, but is presently running only 

a few supplies. The radiological hazard in Building 920 is radiation from activated cooling water and 

oil in transformers (BNL 1997h). 
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Building 928 (Siemens motor generator house). Building 928 is directly adjacent to and at one corner 

connected to Building 929. Both buildings share a basement area. The basement area contains 

support electrical equipment and lubricating equipment for the motor generator set. Also in the 

basement is a mechanical equipment room with water-cooling equipment for power supplies in 

Buildings 928 and 929 and for systems in the AGS ring. There were potential radiation hazards from 

activated cooling water (BNL 1997h).

Building 949. This new building houses the proton target that produces secondaries for the g-2 ring. 

The water-cooled target is highly activated (>50 rem at contact immediately after the beam has been 

turned off). Building 949 is built into the side of the V-line and has extensive shielding. The beam 

line, which runs between the target building and Building 919, contains water-cooled magnets and 

vacuum pumps. The cooling water is shared with the target so it is activated (BNL 1997h). 

Building 960. The building was the site of a large high-energy physics experimental facility, the Seven 

Foot Bubble Chamber (1972–1979). After the Bubble Chamber was removed from Building 960, the 

facility was used for storage and to test superconducting materials, in conjunction with a small 

cryogenic helium refrigerator (1979–1993). The testing of superconducting materials moved to 

Building 902 in 1993. From 1993 to 1997, the building was used for storage and then was 

abandoned. From a radiological standpoint, the particle beam delivered to the Bubble Chamber was 

very low intensity with respect to causing induced activity in the machine components (BNL 1997i, pp. 

3, 6).

Buildings 912, 919, 975. The AGS Department has used warehouse space throughout the facility and 

the Laboratory. The AGS also stored materials outside, in the “inner Mongolia” area of the AGS, the 

southwest area, and the ISABELLE experimental buildings. One of the major issues confronting AGS 

storage is the storage of activated materials. When a part fails in service in the accelerator tunnels or 

the experimental lines, it is often replaced with a spare. After a radiation cooldown period, the 

component is repaired and stored until needed (BNL 1997h). 

Building 918 (Warehouse). This building is adjacent to Building 912. It was built in 1957 and 1958. It 

was a warehouse from the start and always used for that purpose. It was doubled in size in 1962. 

Activated components are stored in it (BNL 1997h).

Building 196 (Warehouse). This wood-frame building was built in 1942. It was formally abandoned 

and demolished in 1995. The building was used by the AGS Department for storage from 1975 until it 

was abandoned in 1991–1992. The building housed depleted uranium, lead, activated copper 

magnet coils, vacuum pumps with oil, and other activated components (BNL 1997h). 

Building 209 (Warehouse). This warehouse belongs to the Safety and Environmental Protection 

(S&EP) division. The AGS has been storing activated components (e.g., magnets, jacks/stands, and 

coils @ <5 mR) in the building since 1995 (BNL 1997h). 

Building 424 (Warehouse). Building 424 was built in 1942 for Camp Upton as its theater. The AGS 

began storing activated components and power supplies with polychlorinated biphenyl capacitors 

there from 1973 until it collapsed in 1996. After the building collapsed, it was totally demolished (BNL 

1997h). 

Building 178 (Warehouses). This building was built in 1922 as a theater and lecture hall. The AGS 

began storing activated components there in the early 1970s. Because the building was wood frame 

and in a "populated" area, the levels of activation were low and much of the equipment was new 

spares. The AGS equipment was removed and the building was demolished in 1992 (BNL 1997h).
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2.4.8

In 1972 the BLIP (Building 931B) and the CLIF (Building 931A) were constructed on a large earthen 

mound. The CLIF was operated independently by the Chemistry Department until 1977, at which time 

it was incorporated into the BLIP operation. Building 931C was constructed in 1996 as part of a 

substantial facility upgrade.

Brookhaven LINAC Isotope Producer, Building 931

The BLIP was designed for production of radioisotopes by irradiation in the LINAC beam. It consists 

of a main building that houses a "hot cell" over a shaft that descends approximately 30 ft. The shaft 

allows targets containing various substances to be lowered along a track to an area into which a 

beam from the LINAC is conducted. Irradiated materials are withdrawn, transported to, and 

processed in, the Target Processing Lab (TPL; Building 801). 

A smaller structure (Building 931A) houses the control panel for BLIP operations and monitoring, as 

well as a single fume hood and a second LINAC beam access system (currently not in use). 

Building 931C acts primarily as a garage to house the fork-lift truck used to deliver irradiated targets to 

the TPL. 

In the BLIP, targets are irradiated with 200-MeV protons. In this process, high-energy secondary 

neutrons are created. These penetrating neutrons are absorbed in the shielding soil around the main 

BLIP containment tank. This leads to some activation of the shielding soil (BNL 1997j, pp. 3, 4).

Building 946 houses the water-cooling equipment for the BLIP target area. This area was upgraded 

to provide better containment and shielding for this activated water system (BNL 1997h, p. 1). 

2.4.9

The Tandem Van de Graaff is housed in Building 901. It consists of two Van de Graaff machines, 

arranged so they can be used independently or in tandem, each capable of accelerating atomic 

particles to energies of 10 MeV. A special device that changes negative ions to positive ions makes it 

possible to achieve a maximum total acceleration of 30 MeV.

Tandem Van de Graaff, Building 901

Building 901 is divided into laboratories, accelerator vaults, and office space. The building is the 

home for research that includes medical radionuclide research, accelerator target development, 

synthetic organic chemistry of radiotracers, and analysis of a variety of materials. Areas inside the 

building include the cyclotron area, the Radio-Frequency Quadrupole (RFQ)-Drift-Tube-LINACS (DTL)

area, laboratory space associated with the Department of Applied Science (DAS), and a mechanical 

shop area that is currently occupied as a storage area. The west end of the building contains the 

Dynamitron, which has been shut down, and a robot room dedicated to the analysis of biological 

samples. 

Area 1 – Cyclotrons. Two cyclotrons housed in this area are used for the production of 

radiopharmaceuticals. The radiopharmaceuticals are processed in the adjacent hot chemistry 

laboratory suite (Rooms 108 to 112). The cyclotrons are operated remotely from Room 118. These 

machines are used almost exclusively for producing the short-lived, positron-emitting isotopes used in 

the PET program. The radioactivity produced by these cyclotrons is remotely transferred, as liquids or 

gases, from the vaults to the adjacent hot laboratory. The hot laboratory contains preparation and 

purification equipment as well as shielded hoods for the processing of the radioisotopes from the 

cyclotrons into radiopharmaceuticals (BNL 1997k, p. 5). 

Area 2 -RFQ-DTL Accelerator. The BNL RFQ-DTL facility consists of the AccSys RFQ accelerator, 

associated electronics, and other support systems. The areas associated with the accelerator include 
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the vault (Room 130), where the accelerator is located, and the control room (Room 129), where the 

power supplies and control console are located. Since 1996, this facility has occupied space formally 

used for the 3.5-MeV Van de Graaff (BNL 1997k, pp. 4, 5).

Area 3 - Laboratory Space. This area is laboratory and office space devoted to DAS projects. Most 

of the space is used for data analysis and storage of materials to be analyzed at the NSLS. The 

former machine shop is used for storage of samples and unused electronic equipment associated with 

the 3.5-MeV Van de Graaff (BNL 1997k, p. 5).

Area 4 - 901W Dynamitron and Robot Room

Most of Building 901 was constructed in 1948 and 1949. It was originally constructed to house the 

cyclotron(s) and the 3.5-MeV Van de Graaff. The 3.5-MeV Van de Graaff was installed in the building 

in 1950. 

. This room in the southwestern corner of Building 901W 

is dedicated to the operation of a robot used by the BNL PET program to analyze plasma samples 

taken from subjects during the PET study. The rest of the area in Building 901W is occupied by the 

Dynamitron (BNL 1997k, p. 5).

In 1958, Building 901 W was constructed to house the 18-in. cyclotron, which, in 1959, was moved 

from the test shack in the rear of the Cosmotron and reassembled in the new building. The 18-in.

cyclotron was capable of accelerating protons to energies of about 3 MeV. From 1960 to 1963, this 

accelerator was used to study elastic and inelastic scattering of polarized neutrons. The target 

material for the production of the neutrons was a T(p,n) 3

He reaction on a target of zirconium tritide. 

The accelerator was capable of consistent beam currents of 100 μA and occasionally reached beam 

currents of 200 μA. 

The 60-in. cyclotron was installed in 1950 and became operational in 1951. The 60-in. cyclotron was 

capable of producing 10 MeV protons, 20 MeV deuterons, and 40 MeV alpha particles. It was used to 

produce a variety of radioisotopes for use in medicine, biology, chemistry and physics. Upgrading of 

the 60-in. cyclotron began in 1964 and was completed in 1968. The upgrade project resulted in the 

cyclotron becoming a variable-energy machine capable of producing 35-MeV protons, 23-MeV 

deuterons, 57-MeV 3

He and 46-MeV alpha particles. During the late 1950s and early 1960s, a 

number of radioisotopes were in development and/or production at what was known then as the Low 

Energy Accelerator Facility (LEAF), which included the cyclotrons, and the Van de Graaff accelerator. 

The isotopes were: 90Y, 47Ca, 67Cu, 38Ar, 28Mg, 124I, 133I, 47Sc, 210Po, 83mKr, 68Ga, 151Sm, 132Te132I 

generator, and 99Mo99mTc generator. The major purpose of these isotopes was medical research. 

In 1965, the Dynamitron was installed in Building 901W. It was used as a positron accelerator until 

1989 to study positron reactions. During this time, it used a large 22Na source as the generator of 

positrons. This source was removed in 1996. At present, the facility is not in operation and there are 

no immediate plans for a restart. All activated items have been removed.

The Van de Graaff was installed in Building 901 in 1950. There were some problems during the early 

years and a reconstruction project was completed in 1954. In 1960 the first experiments with a 3

He 

beam were begun. The first test of the triton beam was carried out in June 1968. During 1968 and for 

a few years after, 28Mg was being produced routinely on the Van de Graaff using the 26 Mg(t,p)28 Mg 

nuclear reaction. During the period from 1968 through the late 1970s, the 3.5-MeV Van de Graaff 

was being used to accelerate tritium for production of isotopes as well as basic physics studies. In 

1985, the Van de Graaff was turned over to the DAS. Some experiments with tritium continued until 

the shutdown of the Van de Graaff in 1991 (BNL 1997k, pp. 6, 7).
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2.4.10

Building 725 is the NSLS. It is one of the most intense sources of X-ray, ultraviolet, and infrared 

radiation in the world. The beams of radiation are created by electrons circulating in two storage 

rings. The X-ray ring energy is 2.5 GeV and the VUV ring is 800 MeV. The radiation produced by the 

circulating electrons travel down 85 independent beamlines to targets on a variety of experiments. 

There are also wet laboratories available for the experimenters, storage space, offices, and a user

machine shop in the building. 

National Synchrotron Light Source, Building 725

The sole occupant of Building 725, which was constructed in 1981, has been NSLS. In 1987, a major 

Phase II expansion of the facility (office space, beamlines, laboratory and setup space) was added, 

but the original function of the facility remained the same. Operations of the VUV ring commenced in 

1982 and the X-ray ring in 1984 (BNL 1997l, pp. 1, 2).

Building 729 is the NSLS Source Development Laboratory. The building houses a linear accelerator 

coupled to a magnetic "wiggler" that produces an intense X-ray beam (BNL 1997m, p. 1). 

2.4.11

The RHIC Complex is shown in Figure 2-4. It is a high-energy particle accelerator that was 

commissioned in 1999 and achieved its first successful operation in the summer of 2000. The RHIC 

is designed to achieve much higher reaction energies by colliding two accelerated beams head-on. 

When the ion beam is traveling at top speed in the AGS, it is taken down another beam line called the 

AGS-To-RHIC (ATR) transfer line. At the end of this line, there is a “fork in the road,” where a 

switching magnet sends the ion bunches down one of two beamlines. Bunches are directed either left 

to the clockwise RHIC ring or right to travel counter-clockwise in the second RHIC ring. From here 

on, the counter-rotating beams are accelerated, as in the Booster and AGS, and then circulate in 

RHIC where they will collide into one another at as many as six interaction points. These intersection 

points are described in Figure 2-5.

Relativistic Heavy Ion Collider Complex, Building 1000 Series

The chronology of the RHIC buildings (BNL 1997n, pp. 4–8):

1981–present Experimental hall, assigned to BRAHMS experiment

Building 1002 

1982–1992 AGS and ADD storage, some activated components

1990–1992 AGS experiment 813

1981–present Utilities, no usage before 1992 

Building 1004A (Service Building)

1992–present RF System testing

1981–present Future experimental area

Building 1004 Open Area

1994–present RF cavities testing for Collider

1996 -present Temporary beamstop for Sextant Test completed during January 1997

1994–present Power supplies, Cryogenic Valve Boxes, other system electronics

Building 1004B (Support Building)

1981–present Commissioned in January 1997 as part of the Sextant Test 

Building 1005 Beam Tunnel
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 Figure 2-4. The Relativistic Heavy Ion Collider complex.

1981-present Tech Shops on 1st Floor, Administrative Offices 2nd, 3rd, 4th Floors

Building 1005S

1991–present STAR Detector

Building 1006

1985–1986 Machining of low-level radioactive steel for AGS Experiment 787

1981–1990 Other short-term uses, temporary storage

1981–present Experimental support, tech shop, utilities, and center room occasional storage

Building 1006A (Service Building)

1981–present Beam tunnel to be commissioned in 1999

Building 1007

1981–1995 Part of tunnel used as a calibration laboratory for Survey Group
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Figure 2-5. RHIC ring intersection points (BNL 2009b).

1991–present PHENIX Experiment

Building 1008

1992–1995 Injection magnet fabrication

1988–1991 Uranium plate machining for Experiment DO at FNAL

1981–present Experimental Support Building, Utilities

Building 1008A

1992–1995 Injection Magnet Group Shop/South Room

1990–1992 AGS/Physics Experiment Prep/South Room

1986–1987 Assembly of Wire Chambers for AGS Experiment 755

1985–1986 Cleaning of Depleted Uranium with TCA for CERN Experiment NA-34

1994–present PHOBOS Experimental

Building 1010

1994–present Future Experimental Hall

Building 1012

2.4.12

Building 801 has from its inception in 1949 been known as the Hot Lab. It consisted of a central 

laboratory, a fanhouse, a radioactive-liquid waste tank farm, and a liquid radioactive-waste 

concentration plant. The original purpose of the central facility was to provide appropriately shielded 

areas for research with large amounts of radioactive material. The building was designed as two 

separate but connected structures; a "Hot" (the western portion) and a "Cold" (the eastern) side. The 

Hot area of the Hot Laboratory included five hot cells, three chemical-processing hot cells, and three 

Hot Laboratory, Building 801 
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high-level hot cells for handling and processing of relatively high levels of radioactivity in gaseous, 

liquid, or solid form. The cells were maintained at negative pressure with respect to their 

surroundings to minimize the possibility of radioactive releases to the building. Each has individual 

exhaust air filters as well as a backup filter preceding discharge to the BGRR stack.

Currently in the Cold area there are a number of chemistry laboratories that conduct basic research. 

On the Hot side, radioisotope processing and production (related to the BLIP), and 

radiopharmaceutical research is conducted by Medical Department staff.

There are Metallurgy Hot Cells [the Metallurgical Evaluation Lab (MEL)] in the Semi Works Area (Area 

56) that were constructed in 1958. Metallurgical studies on highly radioactive reactor components 

were performed in this facility by Department of Applied Technology (DAT) personnel.

There are "D" and "F" waste systems (storage tanks and piping) in the building basement. The D 

waste system holds radioactive nonhazardous liquid waste. The "F" waste system holds 

nonradioactive, nonhazardous liquids. The Hazardous Waste Management Group of S&EP manages 

that facility.

Some additional special facilities in the building are the TPL (a series of six hot boxes in a separate, 

rad-monitored area (Room 66), the MEL, and three inactive Hot Cells. The 300-ft stack south of the 

building receives the ventilation from hoods on the hot side. The stack is monitored for radioactivity

(BNL 1997o, p. 4). 

There was a 60Co irradiation pool in what is now the TPL (Laboratory 66). The pool, which was 

constructed of concrete, had depths of 4, 10, and 14 ft where large amounts of sealed 60Co sources 

were stored under water. There was a drain to D waste in the deep portion of the pool. The pool was 

in operation from about 1952 until about 1960. The 60Co pool was moved, water was drained,

concrete was decontaminated, and the pool area was backfilled with sand (BNL 1997o, p. 7).

From March 1952 to June 1960, there was a program at the Hot Laboratory to produce 131I by aciddissolution of irradiated uranium samples. A separate exhaust system, which included NaOH 

scrubbers and backup charcoal filters, was installed in the hot cell and ducted to a stainless-steel pipe 

in the BGRR stack.

There are pneumatic tubes in the roadway connecting the BGRR to Building 801. These lines 

allowed quick transfer of short-lived isotopes from the reactor to Building 801 where they would be 

used. An interviewee reported that at times the rabbits containing the radioactive samples would 

break or get stuck in these tubes. 

Incidents

On May 15, 1957, a serious incident occurred involving the reprocessing of uranium for the Volatility 

Project. UF6/BrF3 exploded and a few people were injured, one hospitalized. There was a release of 

about 13 lb of unirradiated uranium to the local environment. Damage occurred to nearby trees, 

automobiles, and building equipment. Environmental sampling of the appropriate areas might identify 

residual uranium contamination (BNL 1997o, p. 5).

In August 1960, some highly radioactive waste (reactor fuel elements) was dissolved in aqua regia 

and stored in plastic containers in the storage vault (Room 51). These containers cracked and leaked 

onto a stainless-steel tray containing the plastic containers. The solution was sucked up using an 

aspirator connected to the house vacuum system. The vacuum system became highly radioactive. 

The main vacuum chamber removed all standing liquid in the lines. However, because of the high 

contamination level and long half-life of the activity, the pipes remain residually contaminated. The 

system was moved to a remote location, but some of the piping in the floor is still contaminated as a 
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result of this incident. The piping is still in the concrete floor, which is labeled as a radioactive material 

area (BNL 1997o, p. 7). 

2.4.13

Building 830 is an office and laboratory complex presently occupied by DAT and RHIC personnel. All 

of the RHIC personnel are housed in a modular structure, which is totally office space. The laboratory 

complex has always been used as an extensive experimental area, especially with regard to the use 

of radioactive and hazardous chemicals. This facility houses the Gamma Irradiation Facility (GIF), 

two inactive hot cells, analytical laboratories, electron microscopy, and associated offices. 

High Intensity Radiation Development Laboratory, Building 830

In 1963, Building 830 commenced operations as the HIRDL. At that time, it consisted of the high bay 

area (which houses the two hot cells), laboratories, and offices. Cell 1 (Preparation Cell) was used to 

fabricate high-intensity 60Co sources for food irradiation programs. Some of the sources were then 

moved to Cell 2 (Irradiation Cell) by means of a transfer canal. In the Irradiation Cell, the sources 

were stored in a 21-ft–deep, water-filled pit in the cell until needed. Items to be irradiated were moved 

in and out of the Irradiation Cell by means of two transfer tunnels under the cell. Sources were also 

kept in the two bays for storage and irradiation purposes. 

In 1970, the Low Dosimetry Facility (currently known as the GIF) was added to the northeast end of 

the main building. This facility included the gamma irradiation pool and a machine shop. Sources 

were stored at the bottom of the pool and samples to be irradiated were lowered through air-filled 

tubes. In 1974, the last of the sources was removed from the pit in the Irradiation Cell. Since that 

time, the cells have been used for experiments with lower activity level materials, and storage of 

radioactive materials and contaminated equipment.

At present, the two hot cells in Building 830 have no mission-essential work. The Laboratory spaces 

are being used for experiments, and the GIF currently contains approximately 35,000 Ci of 60Co (BNL 

1997p, pp. 2-3).

2.4.14

Building 555 is the home of BNL's Chemistry Department. It was built between 1963 and 1966. The 

only major modification was made in 1996 when the underground Center for Radiation Chemistry 

Research was added. The building contains three floors plus a partial basement and loft. There are 

many extra-large laboratories with at least two fume hoods per laboratory. There are pipe chases 

between rows of laboratories to deliver all services to the laboratories. Transparent glass piping in the 

basement carries wastewater from the sinks in the laboratories to the sewer system. Counting rooms 

constructed of specially selected low-activity materials are used for low background counting. There 

is a large chemical stockroom with separate storage rooms for acids, bases, and organic solvents, as 

well as a room devoted to the washing of laboratory glassware.

Chemistry Department, Building 555

A large variety of research programs have been conducted during the past 31 years. Many different 

chemicals have been used, including inorganic acids and bases, inorganic compounds and metals, as 

well as a plethora of organic and organometallic compounds. A variety of specialized equipment and 

instrumentation is in use in different laboratories (e.g., there are several laser-based systems in 

operation).

In addition, research with radioactive isotopes has been done since the earliest days. Studies include 

chemical kinetics, hot-atom effects, nuclear chemical studies of nuclear decay schemes, nuclear 

reaction cross-sections, solar neutrino studies, and radiocarbon dating. Over the years, the mix of 

research projects in the Department has changed considerably, but virtually every chemical element 

in the Periodic Table has been used at one time or another in the building (BNL 1997q, p. 5).
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Some radiological problems associated with the building are:

Disposition of sealed 60Co irradiation sources in Laboratories 171–173. These sources are in 

stainless-steel containers that penetrate the laboratory floors and are in the ground below. There is 

evidence that one source (40-Ci source strength) was leaking small amounts of 60Co into the water 

that surrounds the source to provide radiological shielding. This source has been removed and 

disposed of. Some of this contaminated water has gotten into the soil surrounding the buried 

containment vessel. The amount of 60Co in the water and the soil is small. (The estimated 60Co 

activity remaining in this soil is less than 35.) 

Radioactive materials and counting samples are used and/or stored in many rooms: Laboratories 

225, 227, 229, 233, 235, 236, 253, 367, 369, and 374 and counting rooms 103, 203, 204, and 205. 

Periodically, chemical manipulations are done with these materials in vented hoods that are posted as 

radioactive work areas in accordance with BNL procedures. Occurrence reports from 1990 to present 

were reviewed for Building 555. They indicated three separate reportable incidents involving small 

amounts of radioactivity. Two of these, one involving 32P and the other, 238U, were confined to the 

chemistry laboratories in which they occurred. The third involved a Kimwipe contaminated with 124I

that was inadvertently thrown into the trash and was subsequently detected in the garbage truck 

before it left the BNL site (BNL 1997q, p. 9). 

2.4.15

Building 510 was built in 1962 to house the Physics Department personnel and laboratories. It 

consists of approximately 400 offices and 100 laboratories in an area of about 200,000 ft2

. There is a 

large machine shop to support the research programs as well as a number of smaller machine shops. 

A high bay area is available for assembling equipment for experiments. The basement is used 

predominantly for storage but houses the building support facilities such as heating and airconditioning units and power distribution systems. The northeast parking lot contains 11 trailers 

permanently placed for storage of solid materials. These are not climate controlled and no chemical 

or radioactive materials are stored there (BNL 1997r, p. 6).

Physics Department, Building 510

The northeast corner of the building houses in the Health Physics Section (Room 1-136) where there 

are 35 vaults for storage of radioactive materials. There are nine Type A vaults that are 12 in. in 

diameter and penetrate 10 ft into the ground, six Type B vaults that are 8 in. in diameter and penetrate 

10 ft into the ground, and 20 Type C vaults that are 8 in. in diameter and penetrate 3 ft into the 

ground. Vault 17 is contaminated because it contained a leaking radium source. The source has 

been placed in a sealed polyvinyl chloride pipe and is now stored in Vault 4 (BNL 1997r, pp. 4, 5).

2.4.16

2.4.16.1 LMFR Support, Building 820

Other Miscellaneous Radiological Research Facilities

Research and development on Liquid Metal Fuel Reactor (LMFR) technology took place in Building 

820 from 1957 to 1975. A simulated reactor with a core of uranium dissolved in bismuth was studied

(BNL 1997s, p. 4). 

2.4.16.2 Hot Machine Shop, Building 530

Building 530 functioned as a Hot Machine Shop for the Central Shops Division. In 1975, Central 

Shops evacuated the building and relocated its equipment and function to Building 462. The building 

was then used as a storage area by Plant Engineering until 1988, at which time it was demolished

(BNL 1997t, p. 3). 
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2.4.16.3 Health Physics and Safety Instrumentation, Building 535

Building 535 was constructed in 1962 and 1963 and was operational in 1964. The building is a 

multitenant, multifunctional facility consisting of a ground floor and a basement. The Instrumentation 

Division, which is responsible for the building, occupies a major part of the ground floor for research 

and development of X-ray and neutron detectors and corresponding electronics. These detectors and 

associated electronics are used at the NSLS, the HFBR, the AGS, and other worldwide scientific 

research facilities. A state-of-the-art electronic printed circuit design and fabrication facility and a 

microelectronics clean room facility are unique to the Laboratory. The Safety and Environmental 

Protection Division’s Analytical Services Laboratory was in the basement of Building 535 until the 

1990s, when it relocated to Building 490. Personnel Monitoring was also in this building until 2002, 

when it relocated to Building 490. Personnel Monitoring processed film badges and TLDs in the 

building (first floor). Also in the building was the equipment needed to operate the external monitoring 

program (X-ray machine, dark room, densitometer, microscopes, etc.).

In 1978, the NSLS Department moved into the rear section of the basement, occupying that space to 

the present. An underground walkway (tunnel) was constructed in 1986 between the basement of 

Building 535 and the main floor of Building 725 (NSLS Department).

In 1985, Instrumentation Division established the Optical Metrology Laboratory in the southeast 

section of the basement to conduct research for synchrotron radiation experiments. In 1987, an office 

module was added to the southwest corner of the building (BNL 1997u, pp. 5, 6).

2.4.17

The WMF was opened in December 1997 for managing the wastes generated from BNL’s research 

and operation activities. The new WMF replaces the original Hazardous Waste Management Facility 

in its entirety and its design ensures that all storage and transfer activities are accomplished inside 

buildings and on paved and curbed areas. A security fence around the cleared area and a berm 8 ft

high around the perimeter of the radioactive waste portion of the facility ensures that incidental 

exposures meet the BNL design criteria of 25 mrem/yr to nonradiation workers.

Waste Management Facility, Radwaste Reclamation Building, Building 865

The WMF includes an Operations Building (Building 860), a Resource Conservation and Recovery 

Act Waste Building (Building 855), a Radwaste (Radioactive Waste) Reclamation Building (Building 

865), and a Mixed Waste Building (Building 870). 

Radwaste Reclamation Building, Building 865. The Reclamation Building is the primary facility for 

radioactive waste handling, size reduction, and repackaging for subsequent offsite disposal for BNL. 

The goal is to size reduce the radioactive waste stream generated at BNL. This building receives bulk 

radioactive waste of various sizes and configurations to be disassembled, decontaminated, sizereduced, and packaged for temporary storage before shipment off site. To achieve the goal of waste 

reduction, Building 865 will house a 500–lb/d lead smelter, a trash compactor, and a shredder (BNL 

1997v, pp. 2, 3).

Additional radioactive waste treatment facilities include: 

Wastewater processing began in 1995. The Tritium Evaporator Facility (Building 802B) was 

constructed to reduce the total amount of tritiated water released to the Peconic River. The other 

room contains fans that exhaust Building 801 out the stack.

Building 811 was designed in the late 1940s and constructed in 1950 to support the radiological waste 

stream from the Nuclear Engineering Department and Medical Department in the 801 and 701 

complexes. Liquid waste, which contains residual radioactive material and generated on site, is 
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processed at the Waste Concentration Facility (WCF), Building 811. At the WCF, suspended solids 

are removed from the liquid along with a high percentage of radionuclides using a reverse osmosis 

process. The only radionuclide that is not removed during this process is tritium. The tritiated water 

that remains following the waste concentration process is delivered to the Evaporator, where it is 

converted to steam and released as an airborne effluent.

The Waste Management Incinerator. The Laboratory incinerates certain low-level radioactive wastes 

at the HWMF incinerator, Building 444.

2.4.18

The Waste Decontamination Facility was constructed in 1957 and occupies the western half of 

Building 650. It replaced the yard decontamination operations at Building 446. Decontamination of 

large pieces continued until the 1980s. The process slowed and is now idle.

Radiological Waste Decontamination Facility, Building 650

The facility consists of an office, showers, decontamination showers, airlock, and high bay. The 

decontamination showers were once used for decontamination of equipment. They drain to the 

basement then to the D waste tanks (removed). The high bay houses a lead melter (contaminated 

vent pipe; out of service), two vapor blaster talc/water abrasive blasters (internally contaminated; out

of service), one walk-in shot blaster (internally contaminated; out of service), sample hood, and limited 

storage (one B-25 container holding film badge records at the time of the walkthrough). Already 

removed were two wet decontamination tanks (a 20-ft × 4-ft × 4-ft top tank used for 

acid/base/detergent wash).

Associated with the Waste Decontamination Facility is an outside decontamination pad. From 1957 to 

1981, the 20-ft × 20-ft pad was used for washing off surface contamination from large pieces of 

equipment. Drainage from the pad used to go to the underground storage tanks (now removed). The 

surface of the pad has low-level radiological contamination. About 1994, an asphalt cap was laid by 

the Office of Environmental Restoration (OER) to contain the radiological materials. Soil in the vicinity 

of the pad has been characterized by OER and is in its remedial plan (1993 Survey, 137Cs, 152Eu, 

154Eu, depth of 0.5 ft, exceeding OER cleanup goals). The low level contaminants are fixed in the soil 

and are unlikely to migrate.

From 1959 to 1996, the Plant Engineering portion of the building was used as a laundry facility for 

contaminated and clean laundry. In 1996, laundry services were subcontracted and the facility is now 

being used for storage of custodial supplies. Building 650T has been used as offices since it was first 

acquired in 1994 (BNL 1997w, pp. 4-7).

2.5 ISOTOPES OF CONCERN 

Table 2-2 identifies the potential radiological contaminants associated with the specific nuclear 

operations at BNL in terms of isotopes of concern. Due to the wide variety of research activities in a 

single multiple-laboratory building and a significant number of programs that handle and/or create 

practically all of the isotopes of the elements in the Periodic Table, the isotopes of concern are 

identified forensically from the airborne radionuclide releases by the facility. This method is 

considered adequate for the purpose of this section of the BNL site profile; that is, to provide 

background information only. 

Table 2-2. Area information and parameters.

Section Name Description Period

Building 

No. Radionuclide

2.4.1 BGRR U-fueled, graphite-moderated 

and -reflected reactor

1950–1969 701 Ba-140

Ce-144
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Section Name Description Period

Building 

No. Radionuclide

28 MW Cs-137

La-140

Nb-95

Ru-103

Ru-106

Sr-89

Sr-90

Zr-95

U-235

U-238

U-234

Pu-239

2.4.1 BGRR fuel canal BGRR fuel storage 1950–1972 709 Same as 

above

2.4.1 BGRR exhaust 

fanhouse

Highly contaminated ducting 

and fans

1950–1972 704 Cs-137

Sr-90

2.4.1 BGRR exhaust 

filter house

Contaminated HEPA filter 

elements

1950–unknown 708 Cs-137

Sr-90

2.4.2 HFBR 30-to-60-MW thermal, heavywater-moderated research 

reactor

1965–1999 750 Be-7

Br-77

Br-82

Cs-137

H-3

I-126

Mn-54

Xe-133

Xe-135

2.4.2 HFBR Stack 100-m stack and houses iodine 

filters

1965–1999 705 I-131

I-132

I-133

I-134

I-135

2.4.2 CNF Produce extremely-low-energy 

neutron by moderating neutrons

with liquid hydrogen

1980–1996 751

2.4.3 MRC Medical Department 1959–present 490

2.4.4 BMRR Enriched U-fueled, light-watermoderated and -cooled reactor1959–2000 491 Ar-41

Al-26

As-76

Ba-128

Ba-140

Br-82

Ce-141

Ce-144

Co-60

Fe-59

Hg-203

I-124

I-131

I-133

La-140

Mo-99

Na-24

Sb-122
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 38 of 115

Section Name Description Period

Building 

No. Radionuclide

Sc-46

Se-75

Sr-91

Tc-99m

Ti-44

Xe-133

Xe-135

Zn-65

Zn-69m

2.4.5 PET Imaging 

Laboratory

PET scanners for nuclear 

medicine and neurosciences 

research

1977–present 906 C-11

N-13

O-15

F-18

Ge-68

2.4.6 Cosmotron Proton accelerator 1949–1968 902 C-11

N-13

O-15

2.4.7 AGS A large accelerator capable of 

accelerating protons to energies 

up to 30 GeV and heavy ion 

beams to 15 GeV/amu

1960–present 913 C-11

N-13

O-15

2.4.7 200-MeV LINAC Preaccelerator and linear 

accelerator providing proton 

source

1967–present 930 H-3

2.4.7 LINAC cooling 

support service

Provides water-cooling services 

for the LINAC complex

1967–present 930 MER

2.4.7 AGS Booster 

Tunnel

Accumulator booster ring 1960–present 942

2.4.7 AGS Experiment 

Area

Experimental target area for the 

AGS ring and later contains a 

proton beam switchyard system

1960–present 912 Radiation 

sources and 

activated 

targets

2.4.7 AGS ventilation 

system fanhouses

Ventilation exhaust fans 1960–present 913A-E

2.4.7 AGS LINAC Contains 50-MeV AGS LINAC 1960s 914 H-3

2.4.7 E10 Power 

Supply Building

Houses an assortment of power 

supplies

1971–present 920

2.4.7 Siemens motor 

generator house

Electrical equipment and 

lubricating equipment for the 

motor generator

1960–present 928, 929

2.4.7 AGS Department Office, field testing shops, 

machine shops, high bays.

1960–present 911 Radiation 

sources and 

activated 

targets

2.4.7 g-2/Bubble 

Chamber

Contains the 80-in. Bubble 

Chamber

1962–1978 919 H-3

2.4.7 Proton Target Proton target that produces 

secondaries for the g-2 ring

1962–present 949 Ni-63

2.4.7 7-ft Bubble 

Chamber 

High-energy physics 

experimental facility

1972–1979 960 H-3

2.4.7 AGS Warehouses Warehouses 1960–present 975,918,1

96,209,42

4,178

2.4.8 BLIP Production of radioisotopes by 1961–1962 931B As-72
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 39 of 115

Section Name Description Period

Building 

No. Radionuclide

irradiation in the LINAC beam 

(200-MeV protons)

C-11, F-18, N-13 and O-18

As-74

Be-7

Co-57

Co-58

Co-60

Cs-132

Cs-137

H-3

Mn-54

Na-22

O-15

Sc-46

2.4.9 Tandem Van de 

Graaff 

Van de Graaff accelerators

60-in. proton Cyclotron 

producing 10-MeV protons, 20-

MeV deuterons, and 40-MeV 

alpha particles for medical 

radionuclide research

1970–present 901 Yt-90

Ca-47

Cu-67

Ar-38

Mg-28

I-124

I-133

Sc-47

Po-210

Kr-83m

Ga-68

Sm-151

Te-132

I-132

Na-22

H-3

Mo-99

Tc-99m

2.4.10 NSLS Produce intense sources of Xray, ultraviolet, and infrared 

radiation

1981–present 725

2.4.10 NSLS 

Development 

Laboratory

Houses a linear accelerator 1981–present 729

2.2.11 RHIC Complex High-energy particle accelerator 

and collider

1999–present 1002

BRAHMS experimental hall 1981–present 1004A

RHIC service building 1981–present 1004

Open area (future experimental 

area)

1996–present 1004B

RHIC support building 1994–present 1005

Beam tunnel 1981–present 1005S

Offices and tech shops 1981–present 1006

STAR Detector 1985–present 1006A

Service building 1981–present 1007

Beam tunnel 1981–present 1008

PHENIX experiment 1988–present 1008A

Experimental support building 1985–present 1010

PHOBOS experiment 1994–present 1012

Future experimental hall 1994–present

2.4.12 Hot Laboratory Target Processing Laboratory 1951–present 801 As-72

As-74
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Section Name Description Period

Building 

No. Radionuclide

Be-7

Co-57

Co-58

Co-60

Cs-132

Cs-137

H-3

Mn-54

Na-22

Au-199

Bi-213

Br-77

Br-82

Se-75

V-48

I-126

Se-75

Ge-69

2.4.13 HIRDL Radioactive sources fabrication, 

irradiation cells and since 1970, 

the GIF 

1951–present

GIF 1970–

present

830 Co-60

Cs-137

2.4.14 Chemistry 

Department

Studies from chemical kinetics, 

to hot-atom effects, to nuclear 

chemical studies of nuclear 

decay schemes and nuclear 

reaction cross-sections, to solar 

neutrino studies and 

radiocarbon dating.

1966–present 555 H-3

Co-60

2.4.15 Physics 

Department

Laboratories, office space, a 

machine shop, and basement 

area

1962–present 510

2.4.16.1 LMFR Support 

Facility

Research and development 

work for the LMFR

1957–1975 820

2.4.16.2 Hot Machine 

Shop

Hot Machine Shop 1947–1975 530

2.4.16.3 Instrument 

Division

Research and development of 

X-ray and neutron detectors 

and corresponding electronics, 

which are used at the NSLS, 

the HFBR, the AGS, and other 

worldwide scientific research 

facilities.

1964–present 535

2.4.17 Radwaste 

Reclamation 

Building

Located in the WMF for 

processing of radwaste.

1997–present 865

2.4.17 Tritium 

Evaporator 

Facility

Reduce tritium release by 

evaporation

1995–present 802B I-133

Rb-83

Rb-86

Be-7

Co-56

Co-57

Co-58

Co-60

Cs-137
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 41 of 115

Section Name Description Period

Building 

No. Radionuclide

H-3

Mn-54

Na-22

Se-75

Zn-65

2.4.17 Waste 

Management 

Incinerator

Low-level waste incinerator Torn down 444 H-3

Sr-85

I-125

Co-57

Sc-47

Zn-65

C-14

P-32

S-35

Fe-59

Sn-117m

Sn-113

Cr-51

Be-7

2.4.18 Radiological 

Waste 

Decontamination 

Facility

Decontamination of radiological 

waste; hot laundry services

1959–1996 650

In the absence of measurements or studies, NIOSH guidance requires the use of default solubility 

classes and particle size values from the International Commission on Radiological Protection (ICRP; 

NIOSH 2002, pp. 15, 16).

2.6 MAGNITUDE OF SITE ACTIVITY

Table 2-3 provides a perspective of the magnitude of the nuclear operations at BNL. The Site Profile 

preparation guide (ORAUT 2007) requires the magnitude of the operations be expressed in

radioactivity (Ci) of the isotope of concerns. The same information is also required in Section 5, 

Internal Dosimetry. Therefore, radioactivity values in Section 5 should be used for dose 

reconstruction. In cases where available information does not allow the estimation of the radioactivity 

at the time of the operations, the activity fraction of each isotope is listed. Activity fractions listed in 

the table are the ratios of the activity of the individual isotope released and the total activity released 

of all the isotopes. In cases where numeric data are not available, the table provides the descriptive 

text that gives a perspective on the magnitude of the radiological impact on personnel. 

Table 2-3. Magnitude of site activity.

Section Name Description Period

Building 

No. Radionuclide

2.4.1 BGRR U-fueled, graphitemoderated and -reflected 

reactor

28 MW

1950–1969 701 Ba-140

Ce-144

Cs-137

La-140

Nb-95

Ru-103

Ru-106

Sr-89

Sr-90

Zr-95
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 42 of 115

Section Name Description Period

Building 

No. Radionuclide

U-235

U-238

U-234

Pu-239

2.4.1 BGRR fuel canal BGRR fuel storage 1950–1972 709 Same as 

above

2.4.1 BGRR exhaust 

fanhouse

Highly contaminated 

ducting and fans

1950–1972 704 Cs-137

Sr-90

2.4.1 BGRR exhaust 

filter house

Contaminated HEPA filter 

elements

1950–

unknown

708 Same as 

above

2.4.2 HFBR 30-to-60-MW thermal, 

heavy-water-moderated 

research reactor

1965–1999 750 Be-7

Br-77

Br-82

Cs-137

H-3

I-126

Mn-54

Xe-133

Xe-135

2.4.2 HFBR Stack 100-m stack and houses 

iodine filters

1965–1999 705 I-131

I-132

I-133

I-134

I-135

2.4.2 CNF Produce extremely-lowenergy neutron by 

moderate neutron with 

liquid hydrogen

1980–

unknown

751

2.4.3 MRC Medical Department 1959–present 490

2.4.4 BMRR Enriched U-fueled, lightwater-moderated and 

-cooled reactor

1959–2000 491 Ar-41

Al-26

As-76

Ba-128

Ba-140

Br-82

Ce-141

Ce-144

Co-60

Fe-59

Hg-203

I-124

I-131

I-133

La-140

Mo-99

Na-24

Sb-122

Sc-46

Se-75

Sr-91

Tc-99m

Ti-44

Xe-133

Xe-135
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 43 of 115

Section Name Description Period

Building 

No. Radionuclide

Zn-65

Zn-69m

2.4.5 PET Imaging 

Laboratory

PET scanners for nuclear 

medicine and 

neurosciences researches

1977–present 906 C-11

N-13

F-18

O-15

Ge-68

2.4.6 Cosmotron Proton accelerator 1949–1968 902 C-11

N-13

O-15

2.4.7 AGS A large accelerator capable 

of accelerating protons to 

energies up to 30 GeV and 

heavy ion beams to 15 

GeV/amu

1960–present 913 C-11

N-13

O-15

2.4.7 200-MeV LINAC Preaccelerator and linear 

accelerator providing 

proton source

930

2.4.7 LINAC cooling 

support service

Provides water-cooling 

services for the LINAC 

complex

930 MER

2.4.7 AGS Booster 

Tunnel

Accumulator booster ring 942 C-11

N-13

O-15

2.4.7 AGS Experiment 

Area

Experimental target area 

for the AGS ring and later 

contains a proton beam 

switchyard system

1960–present 912

2.4.7 AGS ventilation 

system fanhouses

Ventilation exhaust fans 1960–present 913A-E

2.4.7 AGS LINAC Contains 50 MeV-AGS 

LINAC 

1960s 914 H-3

2.4.7 E10 Power Supply 

Building

Houses an assortment of 

power supply

1971–present 920

2.4.7 Siemens motor 

generator house

Electrical equipment and 

lubricating equipment for 

the motor generator

1960–present 928, 929

2.4.7 AGS Department Office, field testing shops, 

machine shops, high bays.

1960–present 911

2.4.7 g-2/Bubble 

Chamber

Contains the 80-in. Bubble 

Chamber

1962–1978 919 H-3

2.4.7 Proton Target Proton target that produces 

secondaries for the g-2 ring

1962–present 949

2.4.7 7’ Bubble Chamber High-energy physics 

experimental facility

1972–1979 960

2.4.7 AGS Warehouses Warehouses 1960–present 975, 918, 

196, 209, 

424, 178

2.4.8 BLIP Production of radioisotopes 

by irradiation in the LINAC 

beam (200-MeV protons)

C-11, F-18, N-13 and O-18

1961–1962 931B As-72

As-74

Be-7

Co-57

Co-58

Co-60
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 44 of 115

Section Name Description Period

Building 

No. Radionuclide

Cs-132

Cs-137

H-3

Mn-54

Na-22

O-15

Sc-46

Co-56

Ga-68

Ge-69

I-126

Xe-127

Zn-65

2.4.9 Tandem Van de 

Graaff 

Van de Graaff accelerators

60-in. proton Cyclotron 

producing 10-MeV protons, 

20-MeV deuterons, and 40-

MeV alpha particles for 

medical radionuclide 

research

1970–present 901 Sr-90/Y-90

Ca-47

Cu-67

Ar-38

Mg-28

I-124

I-133

Sc-47

Po-210

Kr-83m

Ga-68

Sm-151

Te-132

I-132

Na-22

H-3

Mo-99

Te-99m

2.4.10 NSLS Produce intense sources of 

X-ray, ultraviolet, and 

infrared radiation

1981–present 725

2.4.10 NSLS 

Development 

Laboratory

Houses a linear accelerator 1981–present 729

2.4.11 RHIC Complex High-energy particle 

accelerator and collider

1999–present

BRAHMS experimental hall 1981–present 1002

RHIC service building 1981–present 1004A

Open area (future 

experimental area)

1996–present 1004

RHIC support building 1994–present 1004B

Beam tunnel 1981–present 1005

Offices and tech shops 1981–present 1005S

STAR Detector 1985–present 1006

Service building 1981–present 1006A

Beam tunnel 1981–present 1007

PHENIX experiment 1988–present 1008

Experimental support 

building

1985–present 1008A

PHOBOS experiment 1994–present 1010

Future experimental hall 1994–present 1012
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 45 of 115

Section Name Description Period

Building 

No. Radionuclide

2.4.12 Hot Laboratory Target Processing 

Laboratory

1951–present 801 As-72

As-74

Be-7

Co-57

Co-58

Co-60

Cs-132

Cs-137

H-3

Mn-54

Na-22

Au-199

Bi-213

Br-77

Br-82

Se-75

V-48

I-126

Se-75

Ge-69

U-natural

2.4.13 HIRDL Radioactive sources 

fabrication, irradiation cells 

and, since 1970, the GIF 

1951–present

GIF 1970–

present

830 Co-60

Cs-137

2.4.14 Chemistry 

Department

Studies from chemical 

kinetics, to hot-atom 

effects, to nuclear chemical 

studies of nuclear decay 

schemes and nuclear 

reaction cross-sections, to 

solar neutrino studies and 

radiocarbon dating

1966–present 555

2.4.15 Physics 

Department

Laboratories, office space, 

a machine shop, and 

basement area

1962–present 510

2.4.16.1 LMFR Support 

Facility

Research and development 

work for the LMFR

1957–1975 820

2.4.16.2 Hot Machine Shop Hot Machine Shop 1947–1975 530

2.4.16.3 Instrument Division Research and development 

of X-ray and neutron 

detectors and 

corresponding electronics, 

which are used at the 

NSLS, the HFBR, the AGS, 

and other worldwide 

scientific research facilities.

1964–present 535

2.4.17 Radwaste 

Reclamation 

Building

Located in the WMF for 

processing of radwaste

1997–present 865

2.4.17 Tritium Evaporator 

Facility

Reduce tritium release by 

evaporation

802B I-133

Rb-83

Rb-86

Be-7

Co-56
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 46 of 115

Section Name Description Period

Building 

No. Radionuclide

Co-57

Co-58

Co-60

Cs-137

H-3

Mn-54

Na-22

Se-75

Zn-65

2.4.17 Waste 

Management 

Incinerator

Low-level waste incinerator 444 H-3

Sr-85

I-125

Co-57

Sc-47

Zn-65

C-14

P-32

S-35

Fe-59

Sn-117m

Sn-113

Cr-51

Be-7

2.4.18 Radiological Waste 

Decontamination 

Facility

Decontamination of 

radiological waste

650

2.7 MAJOR INCIDENTS

Table 2-4 describes major site incidents that might have significant potential for internal or external 

exposure to personnel.

Table 2-4. Major incidents.

Facility/building Year Description

BGRR, 

Building 701

1952–57 During the early years of operation of the BGRR, fuel failures occurred that 

resulted in radioactive materials being released to the air stream that cooled 

the reactor. There were 28 reported ruptures of BGRR fuel during the period 

1952–1957. These all occurred with natural uranium. There was one rupture 

of a uranium oxide (U3O8) sample that was being irradiated for the radioiodine

production program. Aside from Ar-41, I-131 was the most important 

radionuclide that would contribute to a potential dose that was discharged to 

the atmosphere from the BGRR. Bromine-82 and I-133 were released in 

somewhat larger concentrations (Meinhold and Meinhold 2001).

Hot Laboratory,

Building 801

5/15/1957 There was a serious incident involving the Volatility Project where uranium 

reprocessing was the objective. There was an explosion of UF6/BrF3 and a 

few people were injured, one hospitalized. There was a release of about 13 lb 

of unirradiated uranium to the local environment. Apparently, the UF6

combines with moisture readily to form an oxide. The BrF3 is very corrosive 

and damaged nearby trees and automobiles and equipment in the building

(BNL 1997o). 

8/1960 There were some highly radioactive waste reactor fuel elements dissolved in 

aqua regia stored in plastic containers in the storage vault (Room 51). These 

containers cracked and leaked onto a stainless-steel tray containing the plastic 

containers. It was decided to suck up the solution with an aspirator connected 
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 47 of 115

Facility/building Year Description

to the house vacuum system. The vacuum system became very highly 

radioactive. The main vacuum chamber removed all standing liquid in the 

lines. However, because of the high level and long half-life of the activity, the 

pipes remain residually contaminated. Thereafter, the system was moved to a 

remote location but some of the piping in the floor is still contaminated as a 

result of this incident. The piping is still in the concrete floor, which is labeled 

as a radioactive area (BNL 1997o). 

HFBR,

Building 750

1997 During a scheduled maintenance shutdown in 1997, a leak in HFBR’s spent 

fuel storage pool was discovered. In November 1999, the Secretary of Energy 

made a decision to permanently close the HFBR (BNL 2001a).

2.8 RADIOLOGICAL ACCESS CONTROLS

Radiological access controls were exercised locally at the site where activities involving radioactive 

materials are performed. Radiological Control Areas are established for each specific laboratory, 

facility, or building in which nuclear operations were conducted at the BNL site. Entry to those areas 

was controlled (i.e., hand-and-shoe counters, dosimeters, and training requirements).
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 48 of 115

3.0

3.1 PURPOSE

OCCUPATIONAL MEDICAL DOSE

AEC-funded work at BNL began in 1948. BNL occupational medical physicians prescribed chest Xrays as part of the preemployment and periodical physical examination of employees from the very 

beginning of operations (Sunderman 1947b; Cowan 1948). These radiographs caused exposure of 

the lungs and other organs and tissues of the body. Exposure of the radiographed workers came 

primarily from the primary X-ray beam and from scattered radiation. 

The purpose of this section is to describe the occupational medical X-ray screening program over time 

at BNL, and to provide organ and tissue doses associated with these exposures. 

3.2 SCOPE

Substantial information regarding the historical BNL occupational medical X-ray program has been 

located. The current medical director and two BNL X-ray technicians were interviewed and some 

older X-ray films were reviewed to gain insight into the occupational medical program and collimation 

practices (Morris 2006a,b,c,d). A peer-reviewed publication (Handloser and Love 1951) provided 

exposure data specific to the BNL radiographic equipment and techniques in use in 1950. A transcript 

of interviews conducted in preparation for an epidemiology study (Brodsky 1964) included discussions 

with BNL‘s first medical director regarding the early radiographic examination practices, equipment,

and exposure rates.

3.3 EXAMINATION FREQUENCY

The occupational medical program at BNL began in 1947. Sunderman (1947b) mentions that the 

recommended X-ray screening protocol should include a preemployment fluororoentgenogram 

(otherwise known as photofluorography or PFG), a preemployment anterior-posterior (AP) and LAT 

lumbar spine, and an X-ray of one forearm. The recommended annual physical should include a 

PFG; and the termination examination should include a PFG, AP and LAT lumbar spine, and the 

same forearm radiographed on the preemployment examination. It was also recommended that 

beryllium workers under contract with the AEC receive a chest X-ray at preemployment, 6-month 

intervals, and at termination. These recommendations specifically referred to chest X-rays on 14-in. ×

17-in. film (Hardy 1948).

The X-ray records in the claim files submitted by BNL do not bear out the recommendations specified 

by Sunderman (1947b). While there are some PFGs in the claim files, there are very few lumbar 

spine examinations, and no X-rays of forearms were found in the sample of cases reviewed [2].

Because no lumbar spine or forearm X-rays were found in claim files records in this early period, it is 

assumed that BNL occupational medical physicians did not implement the protocol recommended by 

Sunderman [3]. 

It is also clear from the claim file records that PFG was not used exclusively in the early years of BNL. 

Beginning in 1947 and continuing until at least 1964, all workers, including pile (reactor) operators, 

were subjected to a preemployment and an annual physical examination, including a PA chest X-ray 

on 14-in. × 17-in. film, but no X-ray on termination (Love 1957, 1959, 1963, 1964; Bond 1963). No 

PFG records were found in the claim files after 1955 [4].

Dose reconstructors should assign dose according to the number of PFG and 14-in. x 17-in. PA chest 

X-rays in the claim file records if they are provided. If these records are not provided, the dose 

reconstructor should assign dose from a preemployment PFG for start dates between 1947 and 1955, 
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 49 of 115

and annual 14-in. × 17-in. PA chests during this period. The default X-ray examination frequencies 

are found in Table 3-1.

Table 3-1. Default frequency of chest X-rays at BNL.

Period Type of chest X-ray Applicability Frequency

1947–1955 PFG All workers Preemployment

Periodic per the records 

No termination examination

1947–1955 14-in. × 17-in. PA All workers Preemployment if no PFG in the records

Annual 

No termination examination

1956–1978 14-in. × 17-in. PA All workers Preemployment 

Annual

No termination examination

1979–present 14-in. × 17-in. PA and

LAT

All workers Preemployment 

Every 7 years, unless records indicate otherwise 

No termination examination

The screening protocol policy seems to have stayed fairly constant (Flood 1976), although the actual 

interval between “annual” X-ray examinations seems to have increased in the mid-1960s, to an 

average of every 2.5 years (Brodsky 1964, pp. 38–39 and p. 72). It is assumed this examination 

frequency pattern continued through 1978. Dose reconstructors should assign dose according the 

number of 14-in. × 17-in. PA chests in the claim file records if they are provided. If these records are 

not provided, the dose reconstructor should assign dose from a preemployment 14-in. × 17-in. PA 

chest, and annual 14-in. × 17-in. PA chests during this period (1965–1978). The default X-ray 

examination frequencies are found in Table 3-1. In 1978, a Presidential Recommendation to Federal 

agencies advised the discontinuation of routine or screening radiographic examinations, including 

chest X-rays, on patients for whom no prior clinical evaluation is made (Carter, J. 1978). This 

recommendation, at least in part, was incorporated into BNL practice and the average time between 

radiographic examinations increased. Starting with 1979, the chest X-ray reexamination cycle for all 

employees might have decreased to once every 7 years (Morris 2006b). Preemployment 

radiographic examination is assumed to have continued as a condition of employment. At some date 

in the 1980s (probably around 1981), the typical radiographic examination expanded to include both 

PA and LAT chest films (Morris 2006a). For expedience, this is assumed to start in 1979 coinciding 

with the extended reexamination frequency. Current (as of 2009) preemployment X-ray examination 

practice is to offer PA and LAT chest X-ray to all employees and it appears that most new employees 

choose to have that examination. For new employees who test positive to tuberculosis, 

preemployment PA and LAT chest X-ray are required. The time between screening chest X-rays has 

been extended over the years and now occurs on a 7-to-10-year cycle (Morris 2006b,c).

Dose reconstructors should use the number of X-rays in the Energy Employee’s X-ray records if 

provided by DOE for the period from 1979 to the present. If records have not been provided, the 

default X-ray examination frequency from Table 3-1 can be used by dose reconstructors in assigning 

dose. 

3.4 EQUIPMENT AND TECHNIQUES

Sunderman (1947c) states that “arrangements have been made with the Powers X-ray Service for 

chest roentgenograms for each employee attached to our Laboratory.” Powers X-ray Service was a 

well-known provider of PFG services on Long Island. Conrad (1948) mentions the continuation of the 

use of Powers X-ray Service for 1948, and further mentions the exposure as a constant 3,000 mrep. 

This is consistent with the information in a retrospective interview, during which the medical director 

stated each examination caused a 4,000-mR exposure (Brodsky 1964, p. 36). Handloser and Love 

(1951) state that the approximate exposure range from the PFG at BNL was 0.7 to 1.2 R. It is 
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assumed that this value is for a single PFG exposure, while the other exposure values reported by 

Conrad and Brodsky are for stereo exposures. Because these values are consistent with the dose 

values in ORAUT-OTIB-0006 (ORAUT 2005), the dose values from that document will be used for 

dose reconstruction of PFG for BNL workers.

Sunderman (1947a) mentions that a decision was made to purchase Westinghouse X-ray equipment. 

It is assumed this equipment was purchased, because Westinghouse X-ray equipment is referred to 

in the published article by Handloser and Love (1951). 

By 1951, the occupational X-ray program was well established and data regarding the equipment and 

exposure had been published in the Journal of Radiology (Handloser and Love 1951). That paper 

listed X-ray tube data for a PFG unit, a fluoroscopic unit, and a radiographic unit. However, it is 

concluded that only the radiographic unit was used for occupational medical screening examinations 

(Brodsky 1964). The 14-in. × 17-in. PA chest radiographic technique used for employee screening 

was 200 mA and 0.10 second at 72 in. with 0.5 mm Al added filtration in the beam. The nominal tube 

potential used was 72 kV and only the tube potential was adjusted to compensate for patient size 

(Handloser and Love 1951). The half-value layer (HVL) of the X-ray beam is assumed to be 1.5 mm 

Al for an X-ray tube with 0.5 mm Al eq. inherent filtration, 0.5 mm Al added filtration, and operated at 

72 kVp (NCRP 1989, Table B.2).

An interview (Morris 2006a) with a registered X-ray technician who worked at BNL from 1960 to 1993 

revealed that a well-maintained Picker machine was in use for chest exposures in 1960. In 1973, 

phantom measurements were made on the Picker machines for various technique factors (Nelson 

1973). In 1978, more measurements were made on the Picker machines, which provide information 

on the HVL and entrance exposures for this period (Carter, N. 1978).

An interview (Morris 2006b) with a registered X-ray technician who worked at BNL from 1993 to the 

present (March 2006) revealed that the equipment currently in service is a single-phase Picker model 

BCX with a measured HVL of 3.2 mm Al. The unit is automatically collimated. A grid cassette film 

holder is routinely used. This unit has been in service since 1991. The unit is tested routinely by a 

medical physicist in accordance with New York State Department of Health Guidelines. Test records 

estimate tube filtration as 4.06 mm Al and the HVL as 3.3 mm Al (Astarita Associates 1998–2004). 

For a PA chest film, the nominal technique in current use is 110 kVp and 3.2 mAs at 72 in. No 

measurements were found for the LAT chest film technique but the technique is similar to PA, except 

it is performed at 6 mAs. 

A sample of 26 PA chest films spanning 1949 through 1971 was reviewed to determine past beam 

collimation practices [5]. The equipment in use appears to have allowed the technician to manually 

set the collimation. In some films before the early 1970s, collimation is evident. In other films from 

the same period, no evidence of collimation is visible. That does not necessarily mean collimation 

was not used. It might indicate that collimation was established just at the edges of the film cassette 

and was, therefore, not visible on the film, which would be consistent with the practice described by 

the technician employed in that timeframe. The earliest film in this sample with obvious collimation 

was made in 1959 (Morris 2006b). Based on this inspection and the statement of the retired 

technician, it is reasonable to assume that all exposures after 1960 were collimated [6].

3.5 ENTRANCE AIR KERMA IN AIR

BNL medical personnel adopted the unusual (possibly unique) practice of recording the entrance skin 

exposure (ESE) for some examinations. Such information is found on the radiologist’s interpretation 

form “Report of X-ray Examination” that is included in the employee’s medical chart. These ESE data 

were based on measurements made during the annual calibration of the X-ray machine. This practice 

of recording ESE was still occurring in 1964 (Brodksy 1964, p. 67). These contemporary data reflect 
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 51 of 115

the exposure for a specific machine and a nominal, not individual-specific, technique. It is not known 

if this exposure included backscatter, and it does not appear to adjust for the worker’s chest 

thickness. Therefore, although it might be a more accurate assessment of the machine output at the 

time of the exposure, it is not likely to significantly improve the accuracy of the dose reconstruction. 

1947 to 1955 PFG

PFG doses are based on the doses in ORAUT-OTIB-0006 (ORAUT 2005). 

1947 to 1959, 14-in. × 17-in. PA Chest

The entrance air kerma in air for the 14-in. × 17-in. radiographs are based on the techniques in 

Handloser and Love (1951), information in National Council on Radiation Protection and 

Measurements (NCRP) Report 102 (NCRP 1989), and ORAUT-OTIB-0006 (ORAUT 2005). The 

calculation follows:

1. From NCRP Report 102 (NCRP 1989), Table B-3, the air kerma rate for a single-phase X-ray 

machine with 2.5 mm Al eq. total filtration and operated at 75 kVp is 0.45 cGy/100 mAs at 100 

cm.

2. Correcting for total filtration of 1.5 mm Al eq using the information in ORAUT-OTIB-0006 

(ORAUT 2005, pg. 10), the air kerma rate is 0.67 cGy/100 mAs at 100 cm.

3. Correcting for the actual mAs (20 mAs) used at BNL according to Handloser and Love (1951), 

the air kerma is 0.134 cGy at 100 cm.

4. Correcting for a source to skin distance of 154 cm, the entrance air kerma in air is 0.06 cGy. 

This is comparable to the 0.055-R value in Handloser and Love (1951).

1960 to 1978

The entrance air kerma for this period is based on the measurements on X-ray equipment reported by 

Carter, N. (1978). Carter reports the following:

1. Measured entrance skin exposure of 11 mR for the PA chest at 115 kV, 100 mA, 1/30 sec 

(Carter 1978).

2. Measured HVL of approximately 3.0 mm Al eq. (Carter 1978).

1979 to 1986

In 1983, measurements of the X-ray machine in Room 2 were made at 120 kVp and 300 mA at 

various exposure times (Zukas 1983). It is clear from these measurements compared to those made 

in 1978 that a grid was probably introduced because the entrance exposures are higher. The 

entrance exposure measurements range from 25.6 mR to 73.0 mR, without distinction between PA 

and LAT. For dose reconstruction, the entrance exposure for the PA chest for this period will be 

assumed to be 30 mR, and 60 mR for the LAT. 

1987 to 1995 

In 1987, measurements of the X-ray machine in Room 2 were made at 100 kVp, 200 and 300 mA at 

various exposure times (Zukas 1987). The entrance exposures listed are lower than those 

determined in 1983, so possibly there was a change in technique factors and/or screen/film 

combination to produce the lower exposures. The entrance exposures range from 8 mR to 35.6 mR, 

with no indication of PA or LAT. For dose reconstruction, the entrance exposure for the PA chest for 

this period will be assumed to be 15 mR, and 30 mR for the LAT.
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 52 of 115

1996 to present

Between 1996 and 2004, the PA chest ESE from the Picker unit in Room 1 was measured at least 

five times by American Board of Radiology/American Board of Medical Physics-certified radiation 

physics consultants during the course of routine inspection. The technique factors were similar to 

those used for nominal PA and LAT chest examination in current use, namely 110 kVp, 3.2 mAs on 

the large focal spot with a 72-in. source-to-image distance and 23-cm phantom. The ESE ranged 

from 6.7 to 8.1 mR. For 1996 to the present (2010), the ESE for PA chest is rounded up to 10 mR to 

account for the possibility that the actual technique used was set to a slightly higher tube potential. 

No measured ESE is available for the LAT view. Therefore, the ESE for the LAT chest was calculated 

to be 25 mR based on the rule of thumb that the ESE for LAT chest is conservatively estimated to be 

2.5 times the ESE for PA chest (ORAUT 2005, p. 20).

3.6 ORGAN DOSE CALCULATIONS

ICRP Publication 34 Method for collimated exposures

The methodology of ICRP Publication 34 (ICRP 1982) is used to estimate the organ doses when the 

tabulated dose conversion factors (DCFs) for collimated X-ray beams are applicable. The method is 

described in Section 3.6.1, Organ Dose from Collimated Exposure. Modifications to this method to 

account for poorly collimated X-ray beams are described in Section 3.6.2, Organ Dose from Poorly 

Collimate Exposure. Calculation of skin dose is described in Section 3.6.3, Skin Dose.

Organ doses are found in Tables 3-2 and 3-3 for various periods at BNL. In these tables, the 

remainder organs are assigned the same dose as the female lung, which is the organ with the highest 

DCF. 

3.6.1

In ICRP Publication 34 (ICRP 1982) methodology, organ dose (OD) is the product of entrance kerma 

(EK) and a DCF. Entrance kerma is defined in ICRP Publication 34 as “air kerma in air without 

backscatter.” For practical application it is numerically the same as, and can be used interchangeable 

with, ESE (ORAUT 2005, p. 16). For organs other than skin, organ dose is calculated as shown in 

Equation 3-1.

Organ Dose from Collimated Exposure

OD = EK × DCF (3-1)

DCF values are a function of beam quality, expressed as the HVL, and of the X-ray projection and 

organ of interest, assuming that the primary beam is properly collimated.

DCFs for some organs are not tabulated in ICRP Publication 34 (ICRP 1982) but are needed for the 

Interactive RadioEpidemiological Program (IREP). In those cases a dosimetric analogy to a nearby 

organ is assumed and the missing DCF is assigned the same value as the analogue organ. ICRP 

reference organs and their IREP organ analogues have been defined (ORAUT 2005, Table 3-2). For 

some organs ICRP Publication 34 lists gender-based DCFs and, in that situation, the larger of the two 

values is assigned as the DCF to the IREP organ analogues.

3.6.2

Organ doses from poorly collimated radiographs are calculated in a similar manner as those for 

collimated exposures. However, in those cases where the DCFs are not available or are not 

applicable, guidance is provided in ORAUT (2005). Overriding any default guidance is the preference 

for actual measurement data when available (ORAUT 2005, p. 15).

Organ Dose from Poorly Collimated Exposure
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 53 of 115

Some organs, specifically the thyroid, ovaries, uterus, and testes, are not in the primary beam of a 

properly collimated PA chest projection. But if the beam is poorly collimated, those organs might be in 

the primary beam. Therefore, a DCF from a different projection is substituted to avoid 

underestimation of the organ dose. ORAUT (2005, Table 9-11) defines appropriate substitutions. 

These substitutions are incorporated in the data shown in Table 3-2.

3.6.3

Skin doses from PA and LAT chest projections are calculated differently than organ doses depending 

on proximity of predefined areas of skin to the primary beam. The primary beam entrance skin dose 

(ENSD), primary beam exit skin dose (EXSD), entrance skin dose outside but near the primary beam 

(ENSDNPB), exit skin dose outside but near the primary beam (EXSDNPB), and remote skin dose 

(RSD) can be calculated.

Skin Dose

ENSD is the product of ESE and a backscatter factor BF, as shown in Equation 3-2. The BF is from 

Table B-8 of NCRP Report 102 (NCRP 1989). The ENSD applies to all skin surfaces in the beam on 

the entrance side of the body. 

ENSD = ESE × BF (3-2)

EXSD is the ENSD divided by a adjusted absorption factor AFF, which is the absorption factor AF for 

the technique tabulated in Table B.7 of NCRP Report 102 (NCRP 1989) decreased by 10% as shown 

in Equations 3-3 and 3-4, respectively. This dose applies to all skin surfaces in the beam on the exit 

side of the body. 

AFF EXSD = ENSD (3-3)

AFF = AF × 0.9 (3-4)

ENSDNPB is 10% of the ENSD as shown in Equation 3-5. This equation is based on the tabulated 

values in ICRP Publication 34 (ICRP 1982) showing that the dose to the testes is 10% of the dose in 

the primary beam when the testes are located just outside the primary beam.

ENSDNPB = ENSD × 0.1 (3-5)

Similarly, EXSDNPB is 10% of the EXSD as shown in Equation 3-6. 

EXSDNPB = ×0.1 (3-6)

RSD, shown in Equation 3-7, is a function of:

• ENSD,

• Square of the distance from the center of the primary beam to the location of interest on the 

skin, 

• The ratio of scattered-to-incident exposure based on the exposure at 1 m due to 90-degree 

scattering of 70-kVp radiation per Table B-2 of NCRP Report 49 (NCRP 1976),

• Average depth dose (ADD) at 12 cm (mid-chest depth) for the technique tabulated in NCRP 

Report 102 (NCRP 1989, Table B.8), and
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• A bias factor of 1.1, which is favorable to the claimant, allows for 10% uncertainty in tabulated 

values of ADD.

2 1 0.0005 1.1 



 



 = × × × ×

R

RSD ENSD ADD (3-7)

Skin doses from PFG, PA, and LAT chests are found in Table 3-3.

3.7 DOSE RECONSTRUCTION

The medical records provided by DOE might include adequate information to define the date, type, 

and count of X-ray examinations that were administered to the claimant as a condition of employment. 

Only radiographs made as a condition of employment are to be included in dose reconstructions. 

Such exposures are limited to PA and LAT chest films and were required or recommended as part of 

a preemployment examination or a routine reexamination. 

If confusion about the radiographic exposure record exists, consider requesting that the notes on the 

exterior of the envelope(s) containing the worker’s X-ray films be transcribed and provided. These 

notes should give insight to the reason the exposures were made, for example preemployment 

examination, routine surveillance, or diagnosis of injury. Assume that any radiograph that was not a 

PA chest or LAT chest was diagnostic, not to be included in dose reconstruction. If the X-ray 

envelope notes associate the annotation “LMD” or “industrial” with a particular exposure, that means 

the radiograph was a diagnostic exposure (i.e., associated with a workplace injury). LMD radiographs 

are not included in dose reconstruction.

Depending on equipment and technique, the photon energy associated with occupational medical Xray dose is in the 30-to-250-keV energy group. Assignment of all BNL occupational medical doses to 

the 30-to-250-keV energy group is favorable to the claimant and recommended. 

3.8 UNCERTAINTY

ORAUT (2005) analyzed uncertainties in the occupational medical X-ray organ doses. The document 

considered several major sources of uncertainty: measurement errors; variations in applied voltage 

(peak voltage), beam current, and exposure time; and uncertainties due to worker size and 

placement. However, at SRS entrance skin dose measurements were made on nine workers of 

varying chest thicknesses (builds) (Cooley 1967). While Cooley (1967) does not report the measured 

chest thicknesses for these nine workers, the entrance skin doses are reported and reflect the 

increases in exposure needed to radiograph thicker body parts, in this case, chests. The standard 

uncertainty of the range of measurements is 5.6, resulting in an uncertainty of 21% from this source. 

Substituting this value into the calculation for combined uncertainty described in ORAUT-OTIB-0006 

(ORAUT 2005) instead of the 10% value used in that document, the resultant standard uncertainty is 

35% from these five sources. For further conservatism, it may be appropriate to assume that errors 

are all positive and that only +35% should be used as suggested in ORAUT-OTIB-0006 (ORAUT 

2005). ORAUT (2005) assesses the relative error in an individual ESE or organ dose to be ±30% at 1 

standard deviation.

Table 3-2. Organ doses (rem) for PFG, PA, and LAT 14-in. × 17-in. chest radiography.

Organ Projection

1947–1955 

PFGa

1947–1959 

14-in. × 17-in. 1960–1978 1979–1986 1987–1995

1996 to 

present

Thyroid PA 5.22E-01 7.26E-03 5.06E-04 1.38E-03 6.90E-04 6.20E-04

LAT 7.98E-03 3.99E-03 3.78E-03

Eye/brain PA 9.60E-02 7.26E-03 5.06E-04 1.38E-03 6.90E-04 6.20E-04
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Organ Projection

1947–1955 

PFGa

1947–1959 

14-in. × 17-in. 1960–1978 1979–1986 1987–1995

1996 to 

present

LAT 7.98E-03 3.99E-03 3.78E-03

Ovaries PA 2.50E-02 2.50E-02b 1.98E-05 5.40E-05 2.70E-05 3.20E-05

LAT 5.40E-05 2.70E-05 4.00E-05

Urinary/bladder PA 2.50E-02 2.50E-02b 1.98E-05 5.40E-05 2.70E-05 3.20E-05

LAT 5.40E-05 2.70E-05 4.00E-05

Colon/rectum PA 2.50E-02 2.50E-02b 1.98E-05 5.40E-05 2.70E-05 3.20E-05

LAT 5.40E-05 2.70E-05 4.00E-05

Testes PA 5.00E-03 5.00E-03 1.10E-07 3.00E-07 1.50E-07 1.00E-07

LAT 6.00E-06 3.00E-06 2.50E-06

Lungs (male) PA 1.26E+00 1.46E-02 5.46E-03 1.49E-02 7.44E-03 5.65E-03

LAT 1.42E-02 7.08E-03 6.90E-03

Lungs (female) PA 1.35E+00 1.50E-02 5.89E-03 1.61E-02 8.03E-03 6.10E-03

LAT 1.60E-02 8.01E-03 7.75E-03

Thymus PA 1.35E+00 1.50E-02 5.89E-03 1.61E-02 8.03E-03 6.10E-03

LAT 1.60E-02 8.01E-03 7.75E-03

Esophagus PA 1.35E+00 1.50E-02 5.89E-03 1.61E-02 8.03E-03 6.10E-03

LAT 1.60E-02 8.01E-03 7.75E-03

Stomach PA 1.35E+00 1.50E-02 5.89E-03 1.61E-02 8.03E-03 6.10E-03

LAT 1.60E-02 8.01E-03 7.75E-03

Bone surface PA 1.35E+00 1.50E-02 5.89E-03 1.61E-02 8.03E-03 6.10E-03

LAT 1.60E-02 8.01E-03 7.75E-03

Liver/gall 

bladder/ spleen

PA 1.35E+00 1.50E-02 5.89E-03 1.61E-02 8.03E-03 6.10E-03

LAT 1.60E-02 8.01E-03 7.75E-03

Remainder 

organs

PA 1.35E+00 1.50E-02 5.89E-03 1.61E-02 8.03E-03 6.10E-03

LAT 1.60E-02 8.01E-03 7.75E-03

Breast PA 1.47E-01 1.08E-03 7.59E-04 2.07E-03 1.04E-03 9.10E-04

LAT 1.72E-02 8.61E-03 7.90E-03

Uterus PA 2.50E-02 2.50E-02b 2.53E-05 6.90E-05 3.45E-05 3.00E-05

LAT 5.40E-05 2.70E-05 3.50E-05

Bone marrow 

(male)

PA 2.76E-01 2.94E-03 1.29E-03 3.51E-03 1.76E-03 1.46E-03

LAT 2.88E-03 1.44E-03 1.53E-03

Bone marrow 

(female)

PA 2.58E-01 2.58E-03 1.23E-03 3.36E-03 1.68E-03 1.41E-03

LAT 2.28E-03 1.14E-03 1.20E-03

Entrance skinc PA 4.05E+00 7.80E-02 1.54E-02 4.20E-02 2.10E-02 1.40E-02

LAT 8.40E-02 4.20E-02 3.50E-02

a. All organ doses from ORAUT-OTIB-0006 (ORAUT 2005). The doses are for stereo PFG (i.e., two exposures).

b. Based on measured doses from ORAUT-OTIB-0006 (ORAUT 2005).

c. Entrance skin dose is the entrance air kerma in air multiplied by the backscatter factor of 1.35 or 1.4 for HVLs of 1.5 or 

4.0 mm Al, respectively, from NCRP Report 102 (NCRP 1989).

Table 3-3. Skin dose guidance for various chest projections and periods.

Area of skin PFG

<1970 

PA chest

>1970 

PA chest

>1970 

LAT chest

Right front shoulder EXSD EXSD EXSD ENSD

Right back shoulder ENSD ENSD ENSD ENSD

Left front shoulder EXSD EXSD EXSD EXSD

Left back shoulder ENSD ENSD ENSD EXSD

Right upper arm to elbow 10% ENSD ENSD 10% ENSD ENSD

Left upper arm to elbow 10% ENSD ENSD 10% ENSD EXSD

Left hand ENSD ENSD 10% ENSD 10% ENSD

Right hand ENSD ENSD 10% ENSD 10% ENSD

Left elbow, forearm, wrist

Right elbow, forearm, wrist

10% ENSD ENSD 10% ENSD 10% ENSD
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Area of skin PFG

<1970 

PA chest

>1970 

PA chest

>1970 

LAT chest

Right elbow, forearm, wrist

Right elbow, forearm, wrist

10% ENSD ENSD 10% ENSD 10% ENSD

Right side of head (including ear) 10% ENSD 10% ENSD 10% ENSD 10% ENSD

Left side of head (including ear) 10% ENSD 10% ENSD 10% ENSD 10% ENSD

Front left thigh RSD (0.52m) RSD (0.52m) RSD (0.52m) RSD (0.52m)

Back left thigh RSD (0.52m) RSD (0.52m) RSD (0.52m) RSD (0.52m)

Front right thigh RSD (0.52m) RSD (0.52m) RSD (0.52m) RSD (0.52m)

Back right thigh RSD (0.52m) RSD (0.52m) RSD (0.52m) RSD (0.52m)

Left knee and below RSD (0.86m) RSD (0.86m) RSD (0.86m) RSD (0.86m)

Right knee and below RSD (0.86m) RSD (0.86m) RSD (0.86m) RSD (0.86m)

Left side of face Eye/Brain Eye/Brain Eye/Brain 10% ENSD

Right side of face Eye/Brain Eye/Brain Eye/Brain 10% ENSD

Left side of neck 10% ENSD ENSD 10% ENSD 10% ENSD

Right side of neck 10% ENSD ENSD 10% ENSD 10% ENSD

Back of head 10% ENSD 10% ENSD 10% ENSD 10% ENSD

Front of neck Eye/Brain Eye/Brain Thyroid 10% ENSD

Back of neck 10% ENSD ENSD 10% ENSD 10% ENSD

Front torso: base of neck to end of sternum EXSD EXSD EXSD Lung

Front torso: end of sternum to lowest rib EXSD EXSD EXSD Lung

Front torso: lowest rib to iliac crest EXSD EXSD 10% EXSD 10% Lung

Front torso: Iliac crest to pubis 10% EXSD 10% EXSD 10% EXSD 10% Lung

Back torso: base of neck to midback ENSD ENSD ENSD Lung

Back torso: midback to lowest rib ENSD ENSD ENSD Lung

Back torso: lowest rib to iliac crest ENSD ENSD 10% ENSD 10% Lung

Back torso: buttocks (Iliac crest and below) 10% ENSD 10% ENSD 10% ENSD 10% Lung

Right torso: base of neck to end of sternum ENSD ENSD ENSD ENSD

Right torso: end of sternum to lowest rib ENSD ENSD ENSD ENSD

Right torso: lowest rib to iliac crest ENSD ENSD 10% ENSD 10% ENSD

Right torso: Iliac crest to pubis (right hip) 10% ENSD 10% ENSD 10% ENSD 10% ENSD

Left torso: base of neck to end of sternum ENSD ENSD ENSD EXSD

Left torso: end of sternum to lowest rib ENSD ENSD ENSD EXSD

Left torso: lowest rib to iliac crest ENSD ENSD 10% ENSD 10% EXSD

Left torso: iliac crest to pubis (left hip) 10% ENSD 10% ENSD 10% ENSD 10% EXSD
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6/2010 Page 57 of 115

Table 3-4. Skin dose (rem) from various chest projections.

a

Area of skin

PFG 

1947–

1955

PA chest 

1947–

1959

PA chest 

1960–

1978

PA chest 

1979–

1986

LAT chest 

1979–

1986

PA chest 

1987–

1995

LAT chest 

1987–

1995

PA chest 

1996–

present

LAT chest 

1996–

present

Right front shoulder 8.82E-02 6.E-04 4.E-04 1.1E-03 8.40E-02 5.E-04 4.20E-02 4.E-04 3.50E-02

Right back shoulder 4.05E+00 7.80E-02 1.54E-02 4.20E-02 8.40E-02 2.10E-02 4.20E-02 1.40E-02 3.50E-02

Left front shoulder 8.82E-02 6.E-04 4.E-04 1.1E-03 4.E-04 5.E-04 2.E-04 4.E-04 2.E-04

Left back shoulder 4.05E+00 7.80E-02 1.54E-02 4.20E-02 4.E-04 2.10E-02 2.E-04 1.40E-02 2.E-04

Right upper arm to elbow 4.05E-01 7.80E-02 1.5E-03 4.2E-03 8.40E-02 2.1E-03 4.20E-02 1.4E-03 3.50E-02

Left upper arm to elbow 4.05E-01 7.80E-02 1.5E-03 4.2E-03 4.E-04 2.1E-03 2.E-04 1.4E-03 2.E-04

Left hand 4.05E+00 7.80E-02 1.5E-03 4.2E-03 8.4E-03 2.1E-03 4.2E-03 1.4E-03 3.5E-03

Right hand 4.05E+00 7.80E-02 1.5E-03 4.2E-03 8.4E-03 2.1E-03 4.2E-03 1.4E-03 3.5E-03

Left elbow, forearm, wrist

Right elbow, forearm, wrist

4.05E-01 7.80E-02 1.5E-03 4.2E-03 8.4E-03 2.1E-03 4.2E-03 1.4E-03 3.5E-03

Right elbow, forearm, wrist

Right elbow, forearm, wrist

4.05E-01 7.80E-02 1.5E-03 4.2E-03 8.4E-03 2.1E-03 4.2E-03 1.4E-03 3.5E-03

Right side of head (including ear) 4.05E-01 7.8E-03 1.5E-03 4.2E-03 8.4E-03 2.1E-03 4.2E-03 1.4E-03 3.5E-03

Left side of head (including ear) 4.05E-01 7.8E-03 1.5E-03 4.2E-03 8.4E-03 2.1E-03 4.2E-03 1.4E-03 3.5E-03

Front left thigh 1.2E-03 1.E-05 5.E-06 1.E-05 1.E-05 7.E-06 7.E-06 5.E-06 7.E-06

Back left thigh 1.2E-03 1.E-05 5.E-06 1.E-05 1.E-05 7.E-06 7.E-06 5.E-06 7.E-06

Front right thigh 1.2E-03 1.E-05 5.E-06 1.E-05 1.E-05 7.E-06 7.E-06 5.E-06 7.E-06

Back right thigh 1.2E-03 1.E-05 5.E-06 1.E-05 1.E-05 7.E-06 7.E-06 5.E-06 7.E-06

Left knee and below 4E-04 4.E-06 2.E-06 5.E-06 5.E-06 2.E-06 3.E-06 2.E-06 2.E-06

Right knee and below 4E-04 4.E-06 2.E-06 5.E-06 5.E-06 2.E-06 3.E-06 2.E-06 2.E-06

Left side of face 9.60E-02 7.3E-03 5.E-04 1.4E-03 8.4E-03 7.E-04 4.2E-03 6.E-04 3.5E-03

Right side of face 9.60E-02 7.3E-03 5.E-04 1.4E-03 8.4E-03 7.E-04 4.2E-03 6.E-04 3.5E-03

Left side of neck 4.05E-01 7.80E-02 1.5E-03 4.2E-03 8.4E-03 2.1E-03 4.2E-03 1.4E-03 3.5E-03

Right side of neck 4.05E-01 7.80E-02 1.5E-03 4.2E-03 8.4E-03 2.1E-03 4.2E-03 1.4E-03 3.5E-03

Back of head 4.05E-01 7.8E-03 1.5E-03 4.2E-03 8.4E-03 2.1E-03 4.2E-03 1.4E-03 3.5E-03

Front of neck 9.60E-02 7.3E-03 5.E-04 1.4E-03 8.4E-03 7.E-04 4.2E-03 6.E-04 3.5E-03

Back of neck 4.05E-01 7.80E-02 1.5E-03 4.2E-03 8.4E-03 2.1E-03 4.2E-03 1.4E-03 3.5E-03

Front torso: base of neck to end of 

sternum

8.82E-02 6.E-04 4.E-04 1.1E-03 1.60E-02 5.E-04 8.0E-03 4.E-04 7.8E-03

Front torso: end of sternum to lowest rib 8.82E-02 6.E-04 4.E-04 1.1E-03 1.60E-02 5.E-04 8.0E-03 4.E-04 7.8E-03

Front torso: lowest rib to iliac crest 8.82E-02 6.E-04 4.E-05 1.E-04 1.6E-03 5.E-05 8.0E-04 4.E-05 8.E-04

Front torso: iliac crest to pubis 8.8E-03 6.E-05 4.E-05 1.E-04 1.6E-03 5.E-05 8.0E-04 4.E-05 8.E-04

Back torso: base of neck to midback 4.05E+00 7.80E-02 1.54E-02 4.20E-02 1.60E-02 2.10E-02 8.0E-03 1.40E-02 7.8E-03

Back torso: midback to lowest rib 4.05E+00 7.80E-02 1.54E-02 4.20E-02 1.60E-02 2.10E-02 8.0E-03 1.40E-02 7.8E-03

Back torso: lowest rib to iliac crest 4.05E+00 7.80E-02 1.5E-03 4.2E-03 1.6E-03 2.1E-03 8.0E-04 1.4E-03 8.E-04
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Area of skin

PFG 

1947–

1955

PA chest 

1947–

1959

PA chest 

1960–

1978

PA chest 

1979–

1986

LAT chest 

1979–

1986

PA chest 

1987–

1995

LAT chest 

1987–

1995

PA chest 

1996–

present

LAT chest 

1996–

present

Back torso: buttocks (Iliac crest and 

below)

4.05E-01 7.8E-03 1.5E-03 4.2E-03 1.6E-03 2.1E-03 8.0E-04 1.4E-03 8.E-04

Right torso: base of neck to end of 

sternum

4.05E+00 7.80E-02 1.54E-02 4.20E-02 8.40E-02 2.10E-02 4.20E-02 1.40E-02 3.50E-02

Right torso: end of sternum to lowest rib 4.05E+00 7.80E-02 1.54E-02 4.20E-02 8.40E-02 2.10E-02 4.20E-02 1.40E-02 3.50E-02

Right torso: lowest rib to iliac crest 4.05E+00 7.80E-02 1.5E-03 4.2E-03 8.4E-03 2.1E-03 4.2E-03 1.4E-03 3.5E-03

Right torso: iliac crest to pubis (right hip) 4.05E-01 7.8E-03 1.5E-03 4.2E-03 8.4E-03 2.1E-03 4.2E-03 1.4E-03 3.5E-03

Left torso: base of neck to end of 

sternum

4.05E+00 7.80E-02 1.54E-02 4.20E-02 4.E-04 2.10E-02 2.E-04 1.40E-02 2.E-04

Left torso: end of sternum to lowest rib 4.05E+00 7.80E-02 1.54E-02 4.20E-02 4.E-04 2.10E-02 2.E-04 1.40E-02 2.E-04

Left torso: lowest rib to iliac crest 4.05E+00 7.80E-02 1.5E-03 4.2E-03 4.E-05 2.1E-03 2.E-05 1.4E-03 2.E-05

Left torso: Iliac crest to pubis (left hip) 4.05E-01 7.8E-03 1.5E-03 4.2E-03 4.E-05 2.1E-03 2.E-05 1.4E-03 2.E-05

a. Values less than 1 mrem shown to one significant digit.
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4.0

4.1 INTRODUCTION

OCCUPATIONAL ENVIRONMENTAL DOSE

Ambient external gamma radiation outside radiologically controlled areas is the result of ubiquitous 

background radiation, or gamma emissions from stack effluent such as 41Ar, or skyshine due to air 

scatter from an otherwise well-shielded radiation source. 

During the early years (1950 to 1973), BNL used an ion chamber and dynamic capacitor electrometer 

assembly to measure ambient external radiation. The ion chamber/dynamic capacitor electrometer 

assemblies and thermoluminescent dosimeters (TLDs) were used concurrently during 1973 and 1974. 

Information on the specific type of TLD used was not published until 1980 when the use of CaF2:Dy 

TLDs was noted (Naidu and Olmer 1981). The site changed to LiF:Mg Tl technology in 2000 (BNL 

2001a). 

Monitoring results from as many as four onsite stations are available for some years between 1967 

and 1984. If available, these data are included in the average dose estimates. Beginning in 1985, the 

perimeter and onsite monitoring program was expanded to approximately 20 onsite stations and 

steadily increased to more than 50 onsite stations as of 2004 [7].

4.2 PURPOSE

The purpose of this section is to identify the sources and quantify the magnitude of the radiation dose 

received by BNL employees due to ambient radiation on the BNL site.

4.3 SCOPE

This section evaluates the ambient external radiation based on the monitoring results from four 

perimeter stations surrounding the BNL site at the northwest, southwest, southeast, and northeast 

coordinates. 

Doses due to inhalation and ingestion are evaluated using stack emissions and the results of BNL 

environmental reports.

4.4 AMBIENT EXTERNAL RADIATION

4.4.1

Before 1962, average and maximum environmental external gamma doses are derived from summary 

data published in retrospect (Meinhold and Meinhold 2001). During the period from 1962 through 

1966, doses are derived from measurements at three or four continuous onsite monitoring stations as 

published in annual environmental monitoring reports. The standard deviation of the data for these 

years assumes a normal distribution. Table 4-1 provides average and maximum external gamma 

radiation results that have been adjusted to reflect a 2,000 hr/yr occupancy. After 1966, onsite

monitoring results were no longer reported and the only data available are fenceline measurements

that, in general, are remote from areas occupied by employees. The site averages for 1962 through 

1966 have a mean of 71.4 mrem/yr. The maximum value for these years is 141 mrem/yr, which is 

approximately twice the average. The average is exceeded only once during this 5-year period. 

Based on this data, a reasonable average that is favorable to the claimant for years beyond 1966 is 

75 mrem/yr with a maximum of twice the average, suggesting a 100% uncertainty if the average is 

assigned as a normal distribution.

Use of Offsite Monitoring
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4.4.2

For dose reconstruction, the ambient external dose should be applied to all unmonitored workers. For 

monitored workers, external dose should be based on the dosimetry records without assignment of 

ambient external dose.

Applicability for Dose Reconstruction

Table 4-1. External gamma radiation dose (mrem/yr).

Year

Monitoring 

location P-9a

Monitoring 

location 

S-10b

Monitoring 

location 

S-11b

Monitoring 

location 

S-12b

Site 

average

Error 1

standard 

deviationc Maximum

1947d 0 0

1948d 0 0

1949d 0 0

1950e 0.18 0.27

1951e 6.4 9.2

1952e 8.0 11

1953e 8.5 12

1954e 8.9 13

1955e 8.9 13

1956e 8.5 12

1957e 7.3 11

1958e 21 30

1959e 27 39

1960e 32 50

1961e 27 39

1962f 44 35 59 74 53 17 74

1963f 135 141 125 127 132 8 141

1964f 42 64 41 49 13 64

1965f 40 97 38 58 34 97

1966g 40 128 28 65 54 128

1967–

presentg

75 150

a. Monitoring location P-9 was originally known as E-9.

b. Monitoring locations S-10, S-11, and S-12 were originally known as E-10, E-11, and E-12.

c. Error term includes only counting uncertainty and does not include sampling uncertainty.

d. Data before 1950 are assumed based on site activities in those years.

e. Data from 1950 to 1961 are from Meinhold and Meinhold (2001).

f. Data from 1962 to 1965 are extracted from annual environmental monitoring reports (Hull 1963, 1964, 1966; Hull and 

Gilmartin 1967).

g. Data after 1965 are assumed based on personnel dosimetry program criteria.

4.5 INHALATION OF ONSITE AIRBORNE RADIONUCLIDES 

BNL has monitored releases to the environment since 1950 when the first stack became operational 

(Meinhold and Meinhold 2001). Stack release points and principal radionuclides in the airborne 

effluent are contained in the site annual environmental reports.

4.5.1

With the exception of tritium, the intake for each radionuclide is the product of the highest 

Stack Monitoring

annual 

average concentration reported for that radionuclide at any stack or perimeter sampling location 

during the year and the assumed breathing rate of 3.3 × 10-4 m3

/s (ICRP 1975) based on a 

2,000-hr/yr exposure.
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The individual stacks and their period of operation are:

Stack ID Period of operation Stack ID Period of operation

BGRR 1950–1969 Chem 1973–1993

HFBR 1966–2004 BMRR 1996–1997

Hot Lab & HFBR 1986 Incinerator 1981–1996

Hot Lab 1987–2004 Evaporator 1995–2000

BLIP/LINAC 1973–2004

The following isotopes were identified:

H-3 Co-60 Cs-137

Na-22 Zn-65 Ba-140

Na-24 Rh-106 La-140

Ar-41 Ru-106 Ce-141

Mn-54 I-125 Ce-144

Co-58 I-131 Eu-152

Fe-59 Cs-134 Eu-155

4.5.2

The intake of each radionuclide that substantially contributes to personnel dose is tabulated by year. 

In Table 4-2, the maximum annual intakes are provided. These values are then assigned as a 

lognormal distribution with a GSD of 3.

Estimates of Potential Inhalation Intake

As stated above, with the exception of tritium, the intake for each radionuclide is the product of the 

highest annual average concentration reported for that radionuclide at any stack or perimeter 

sampling location during the year and the assumed breathing rate of 3.3 × 10-4 m3

/s (ICRP 1975) 

based on a 2,000-hr/yr exposure. The 3

H intake, assumed to be in the form of water vapor, is 

determined as described for other radionuclides, and then increased by a factor of 1.5 to account for 

direct absorption through the skin [8].

In most years, stack effluent sampling shows that some radionuclides were released but not 

specifically monitored at ground-level site or perimeter sampling points. In those instances, the 

highest annual average stack concentration was reduced by a factor of 0.01 to account for the 

lessened overall intake due to contribution from multiple widely spaced facilities, atmospheric 

dispersion, and building wake, and the result was used as though it were a ground-level site or 

perimeter sample. Most radionuclides released from the stacks were not subsequently detected by 

ground-level monitoring stations. Therefore, a reduction factor could not be calculated from the data, 

so a reduction factor of 0.01 was applied to the stack air concentrations to provide a bounding groundlevel concentration [9].

No measurements of particle size or solubility are available. Solubility and particle size for 

environmental intake calculations should be assumed to be the same as those used in Section 5.0 of 

this document. If no data are available, default solubility classes and particle size values from the 

ICRP should be used (NIOSH 2002, pp. 15–16). 

4.6 INGESTION

Potable water, supplied from onsite wells, provides the most likely pathway for radionuclide ingestion 

at BNL. Environmental reports include the radionuclide concentration in potable water. Most of the 

observed radionuclides are naturally occurring and consistent with concentrations found in offsite 

regional water supplies. 
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Table 4-2. Maximum annual environmental occupational radionuclide inhalation (Bq/yr).

a,b

Year

Radionuclide

H-3 I-131 1-125 Na-22 Na-24 Mn-54 Co-58 Fe-59 Co-60 Zn-65 Sr-90

Nb-95

Zr-95 Tc-99c

1950 3.01E+04 2.70E-01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1951 3.01E+04 4.73E+01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1952 3.01E+04 6.09E+01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1953 3.01E+04 6.42E+01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1954 3.01E+04 6.76E+01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1955 3.01E+04 6.76E+01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1956 3.01E+04 6.42E+01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1957 3.01E+04 5.41E+01 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1958 3.01E+04 6.15E+02 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1959 3.01E+04 6.15E+02 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1960 3.01E+04 6.15E+02 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1961 3.01E+04 6.15E+02 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1962 3.01E+04 7.03E+02 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 2.11E+02 N/A

1963 3.01E+04 6.15E+02 4.84E+04 5.51E-02 1.19E+03 3.78E+01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 8.30E+02 N/A

1964 3.01E+04 5.71E+02 4.84E+04 5.51E-02 1.19E+03 3.52E+00 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 5.27E+00 N/A

1965 3.01E+04 7.45E+02 4.84E+04 5.51E-02 1.19E+03 6.15E-01 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 5.27E+00 N/A

1966 3.01E+04 8.11E+02 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 8.79E-01 N/A

1967 2.11E+05 3.16E+02 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 4.84E+00 N/A

1968 1.71E+06 7.47E+05 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 1.05E-01 3.52E-01 4.84E+00 N/A

1969 1.71E+06 1.32E+01 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 2.64E-01 3.52E-01 6.68E+00 N/A

1970 3.96E+06 2.64E-01 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 7.91E-01 3.52E-01 8.79E+00 N/A

1971 1.05E+03 8.79E-02 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 3.52E-01 1.76E-01 8.09E+00 N/A

1972 7.91E+02 8.79E-02 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 8.79E-02 1.76E-01 1.58E+00 N/A

1973 3.41E+06 8.79E-02 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 8.79E-02 1.76E-01 2.64E-01 N/A

1974 4.46E+06 1.76E-02 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 1.26E+01 1.76E-01 3.08E+00 N/A

1975 3.42E+06 1.76E-02 4.84E+04 5.51E-02 1.19E+03 8.79E-02 3.22E-01 1.22E+04 1.49E-03 8.79E-02 1.76E-01 1.05E+00 N/A

1976 4.08E+06 1.76E-02 4.84E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.22E+04 1.49E-03 8.79E-02 1.76E-01 1.05E+00 N/A

1977 4.11E+03 8.79E-01 4.84E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.22E+04 1.49E-03 8.79E-02 1.76E-01 8.88E+00 N/A

1978 5.32E+05 8.79E-01 4.84E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.22E+04 1.49E-03 8.79E-02 1.76E-01 8.88E+00 N/A

1979 1.37E+06 8.79E+00 4.84E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.22E+04 1.49E-03 3.42E+04 1.76E-01 8.88E+00 N/A

1980 2.05E+06 8.79E+00 4.84E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.22E+04 1.49E-03 3.42E+04 1.76E-01 8.88E+00 N/A

1981 1.32E+07 4.88E+03 4.84E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.22E+04 1.49E-03 3.42E+04 1.76E-01 8.88E+00 4.88E+03

1982 2.05E+07 4.88E+03 6.35E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 4.88E+03 1.49E-03 3.42E+04 1.76E-01 8.88E+00 4.88E+03

1983 1.90E+07 4.88E+03 8.30E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 4.88E+03 1.49E-03 3.42E+04 1.76E-01 3.52E-03 4.88E+03

1984 6.91E+06 1.47E+04 4.88E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 4.88E+03 2.37E-03 3.42E+04 1.76E-01 3.52E-03 4.88E+03

1985 9.66E+05 1.37E+00 2.98E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.47E+02 2.64E-02 1.40E-02 1.76E-01 3.52E-03 4.88E+03

1986 6.92E+06 1.03E+03 2.54E+04 5.51E-02 1.19E+03 4.88E+02 3.22E-01 1.47E+02 1.15E+00 1.40E-02 1.76E-01 3.52E-03 4.88E+03

1987 1.15E+07 8.30E+03 4.35E+03 5.51E-02 1.19E+03 1.70E-03 3.22E-01 4.88E+01 1.15E+00 1.40E-02 1.76E-01 3.52E-03 2.05E+03

1988 1.71E+06 4.88E+02 9.52E+02 5.51E-02 1.19E+03 5.37E-01 3.22E-01 4.88E+01 4.36E-01 3.24E-01 1.76E-01 3.52E-03 2.44E-01
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Year

Radionuclide

H-3 I-131 1-125 Na-22 Na-24 Mn-54 Co-58 Fe-59 Co-60 Zn-65 Sr-90

Nb-95

Zr-95 Tc-99c

1989 2.70E+06 8.88E-02 1.03E+04 1.03E+00 1.19E+03 5.37E-01 3.22E-01 4.88E+01 6.44E-01 8.16E+00 1.76E-01 3.52E-03 2.44E-01

1990 4.63E+05 6.94E+02 2.30E+03 1.03E+00 1.19E+03 3.81E-02 3.22E-01 4.88E+01 6.55E+00 5.37E-03 1.76E-01 3.52E-03 2.44E+02

1991 6.44E+05 6.94E+02 4.75E+03 5.11E+00 1.19E+03 4.75E-02 1.34E+00 4.88E+01 2.83E-01 2.22E+00 1.76E-01 3.52E-03 2.44E+02

1992 4.44E+05 6.94E+02 2.59E+03 5.11E+00 1.19E+03 2.44E+01 1.34E+00 4.88E+01 2.77E+00 2.22E+00 1.76E-01 3.52E-03 2.44E+02

1993 5.33E+05 3.61E-02 2.44E+04 3.11E-01 1.19E+03 7.89E+00 6.83E+00 1.99E+01 6.65E-01 2.22E+00 1.76E-01 3.52E-03 2.44E+02

1994 5.92E+05 3.61E-02 2.44E+04 3.11E-01 1.19E+03 2.19E-01 6.83E+00 1.99E+01 9.49E-04 1.17E-01 1.76E-01 3.52E-03 2.44E+02

1995 7.15E+05 3.21E+00 1.95E+03 2.52E-03 1.19E+03 2.78E-02 6.74E-01 1.99E+01 3.84E-02 1.17E-01 1.76E-01 3.52E-03 2.44E+02

1996 3.52E+05 3.21E+00 1.47E+03 1.12E-01 1.19E+03 4.30E-02 6.74E-01 1.99E+01 4.88E-03 1.01E+02 1.76E-01 3.52E-03 2.44E+02

1997 1.98E+05 1.72E+02 1.47E+03 4.55E-03 1.19E+03 7.42E-02 4.18E-01 1.99E+01 1.37E+01 1.01E+02 1.76E-01 3.52E-03 N/A

1998 2.73E+05 1.72E+02 1.47E+03 6.59E-03 1.19E+03 1.21E-02 1.99E-01 1.99E+01 7.03E-03 3.92E-01 1.76E-01 3.52E-03 N/A

1999 1.33E+05 1.72E+02 7.86E+00 6.59E-03 1.19E+03 9.33E-02 8.84E-02 1.99E+01 1.18E-01 2.17E+00 1.76E-01 3.52E-03 N/A

2000 3.52E+04 1.72E+02 7.86E+00 1.08E-03 1.19E+03 1.53E-02 1.45E-02 1.99E+01 1.93E-02 3.56E-01 1.76E-01 3.52E-03 N/A

2001 2.64E+04 3.83E-05 7.86E+00 1.08E-03 1.19E+03 1.53E-02 1.45E-02 1.99E+01 1.93E-02 3.56E-01 1.76E-01 3.52E-03 N/A

2002 2.37E+04 3.83E-05 7.86E+00 1.08E-03 1.19E+03 1.53E-02 1.45E-02 1.99E+01 1.93E-02 3.56E-01 1.76E-01 3.52E-03 N/A

2003 5.92E+05 3.83E-05 7.86E+00 1.08E-03 1.19E+03 1.53E-02 1.45E-02 1.99E+01 5.25E-08 3.56E-01 1.76E-01 3.52E-03 N/A

2004 5.92E+05 3.83E-05 7.86E+00 1.08E-03 1.19E+03 1.53E-02 1.45E-02 1.99E+01 5.25E-08 3.56E-01 1.76E-01 3.52E-03 N/A

Year

Radionuclide

Tc-99md Ru-103

Rh-106

Ru-106 Cs-134e Cs-137 Ba-140 La-140 Ce-141 Ce-144 Eu-152 Eu-155d Ra-226 Th-228

1950 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1951 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1952 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1953 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1954 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1955 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1956 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1957 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1958 N/A 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 N/A 6.07E-03 2.55E-01 1.76E-01

1959 9.82E+03 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1960 9.82E+03 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1961 9.82E+03 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1962 9.82E+03 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1963 9.82E+03 6.51E+01 5.27E+01 N/A 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1964 9.82E+03 8.79E-01 1.23E+01 N/A 7.03E+00 1.23E+00 1.23E+00 8.79E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1965 9.82E+03 8.79E-01 2.20E+00 N/A 2.37E+00 1.23E+00 1.23E+00 8.79E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1966 9.82E+03 6.15E-01 5.27E-01 N/A 7.03E-01 1.23E+00 1.23E+00 4.40E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1967 9.82E+03 1.05E+00 5.27E-01 N/A 7.03E-01 1.23E+00 1.23E+00 4.40E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1968 9.82E+03 1.05E+00 5.27E-01 N/A 6.15E-01 5.27E-01 5.27E-01 4.40E-01 6.33E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1969 9.82E+03 1.05E+00 2.11E+00 N/A 5.27E-01 1.76E-01 1.76E-01 4.40E-01 1.58E+01 8.76E-07 6.07E-03 2.55E-01 1.76E-01
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Year

Radionuclide

Tc-99md Ru-103

Rh-106

Ru-106 Cs-134e Cs-137 Ba-140 La-140 Ce-141 Ce-144 Eu-152 Eu-155d Ra-226 Th-228

1970 9.82E+03 1.05E+00 2.11E+00 N/A 1.23E+00 1.76E-01 1.76E-01 4.40E-01 2.46E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1971 9.82E+03 1.05E+00 2.11E+00 N/A 1.05E+00 8.79E-02 8.79E-02 4.40E-01 6.15E-01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1972 9.82E+03 1.05E+00 2.11E+00 N/A 4.40E-01 2.64E-01 2.64E-01 4.40E-01 2.37E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1973 9.82E+03 1.05E+00 1.23E+00 1.23E-01 8.79E-02 2.64E-01 2.64E-01 4.40E-01 2.37E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1974 9.82E+03 1.05E+00 1.23E+00 1.23E-01 8.79E-02 2.64E-01 2.64E-01 4.40E-01 4.22E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1975 9.82E+03 1.05E+00 6.15E-01 1.23E-01 8.79E-02 3.52E-01 2.64E-01 4.40E-01 6.68E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1976 9.82E+03 4.88E+03 6.15E-01 1.23E-01 8.79E-02 3.52E-01 3.52E-01 4.40E-01 8.00E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1977 9.82E+03 4.88E+03 7.91E-01 1.23E-01 8.79E-01 3.52E-01 3.52E-01 3.52E-02 1.32E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1978 9.82E+03 4.88E+03 7.91E-01 1.23E-01 8.79E-01 3.52E-01 3.52E-01 3.52E-02 3.52E-01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1979 9.82E+03 4.88E+03 7.91E-01 1.23E-01 8.79E-01 2.20E-01 2.20E-01 3.52E-02 2.02E+00 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1980 9.82E+03 4.88E+03 7.91E-01 1.23E-01 8.79E-01 2.20E-01 2.20E-01 3.52E-02 7.91E-01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1981 9.82E+03 4.88E+03 7.91E-01 1.23E-01 3.96E-01 2.20E-01 2.20E-01 3.52E-02 7.91E-01 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1982 9.82E+03 4.88E+03 9.67E-02 1.23E-01 2.64E+00 2.20E-01 2.20E-01 3.52E-02 5.36E-03 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1983 9.82E+03 4.88E+03 9.67E-02 1.23E-01 2.99E-01 2.20E-01 2.20E-01 3.52E-02 5.36E-03 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1984 9.82E+03 4.88E+03 9.67E-02 1.23E-01 1.77E-01 3.70E-03 2.20E-01 3.52E-02 5.36E-03 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1985 9.82E+03 5.86E+02 9.67E-02 1.23E-01 2.81E-01 3.70E-03 2.20E-01 3.52E-02 5.36E-03 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1986 9.82E+03 5.86E+02 7.03E-02 1.23E-01 2.81E-01 3.70E-03 2.20E-01 3.52E-02 5.36E-03 8.76E-07 6.07E-03 2.55E-01 1.76E-01

1987 4.88E+02 5.86E+02 7.03E-02 9.58E-03 2.17E-01 3.70E-03 2.20E-01 3.52E-02 5.36E-03 8.76E-07 6.07E-03 2.55E-01 5.27E-02

1988 4.88E+02 1.12E+00 7.03E-02 9.58E-03 2.00E-01 3.70E-03 1.86E+02 3.52E-02 5.36E-03 8.76E-07 6.07E-03 8.44E-01 8.79E-02

1989 7.33E+03 1.12E+00 1.28E+02 2.60E-02 4.48E-01 3.70E-03 1.86E+02 3.52E-02 5.36E-03 8.76E-07 8.79E-03 6.15E-01 8.79E-02

1990 7.33E+03 1.52E+01 1.28E+02 2.60E-02 6.15E-01 3.70E-03 1.86E+02 3.52E-02 1.86E+00 8.76E-07 8.79E-03 1.03E+00 3.52E-02

1991 7.33E+03 1.52E+01 1.28E+02 2.60E-02 7.77E-01 3.70E-03 1.86E+02 3.52E-02 1.86E+00 8.76E-07 8.79E-03 1.86E-01 3.52E-02

1992 7.33E+03 1.52E+01 1.28E+02 2.60E-02 2.50E-01 7.76E+01 1.86E+02 9.41E-01 1.86E+00 8.76E-07 8.79E-03 1.86E-01 1.55E-01

1993 7.33E+03 1.52E+01 1.28E+02 2.60E-02 4.55E-01 7.76E+01 1.86E+02 9.41E-01 1.86E+00 8.76E-07 8.79E-03 1.86E-01 1.55E-01

1994 3.03E+02 1.52E+01 1.28E+02 4.49E-03 1.38E+00 7.76E+01 1.86E+02 9.41E-01 1.86E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

1995 3.03E+02 1.52E+01 1.28E+02 4.49E-03 2.76E+01 7.76E+01 1.86E+02 9.41E-01 1.86E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

1996 3.03E+02 1.52E+01 1.28E+02 4.49E-03 5.86E-01 7.76E+01 1.86E+02 9.41E-01 1.86E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

1997 3.03E+02 1.52E+01 1.28E+02 4.49E-03 1.47E-01 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

1998 3.03E+02 1.52E+01 1.28E+02 4.49E-03 1.31E-02 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

1999 3.03E+02 1.52E+01 1.28E+02 4.49E-03 2.59E-01 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

2000 3.03E+02 1.52E+01 1.28E+02 4.49E-03 4.25E-02 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

2001 3.03E+02 1.52E+01 1.28E+02 4.49E-03 4.25E-02 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

2002 3.03E+02 1.52E+01 1.28E+02 4.49E-03 5.44E-04 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

2003 3.03E+02 1.52E+01 1.28E+02 4.49E-03 5.44E-04 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

2004 3.03E+02 1.52E+01 1.28E+02 4.49E-03 1.81E-09 8.07E+02 4.27E+03 9.41E-01 7.34E+00 8.76E-07 8.79E-03 5.07E+00 1.55E-01

a. Highlighted values are bounded by available data. Missing years are filled in with nearest results (highest value if uneven interval). 

b. Values not highlighted are from closest available result.

c. Tc-99 reported only for incinerator, which operated from 1981 to 1996.

d. Tc-99m and Eu-155 are reported for the BMRR, which began operating in 1959; Tc-99m is also reported for incinerator.

e. Cs-134 reported for the BLIP, which began operating in 1973; it is also reported for evaporator.
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 65 of 115

Neither potable groundwater nor soil ingestion has been found to be a pathway of exposure. No 

ingestion dose is indicated [10].

4.7 UNCERTAINTY

A high degree of uncertainty is associated with environmental monitoring. Uncertainty is the result of 

many factors listed here in approximate order of importance: atmospheric dispersion models used to 

describe dilution of radionuclide concentration from the stack to the receptor; location of ground-level 

monitoring stations to represent the unmonitored worker adequately; efficiency of radioactive iodine 

sampling media before use of charcoal collectors; average water intake; occupancy factors for 

workers on the site; average breathing rate; measurement detection limits; and accuracy of 

measurement results. Additional factors could be listed. The error associated with each factor is 

unknown, but can be reasonably expected to range from more than a factor of 10 for atmospheric 

dilution, to about 20% for accuracy of measurement [11].

Although the historical uncertainty cannot be substantially reduced without a great deal of effort, 

importance of uncertainty in dose reconstruction is reduced by the fact that many of the doses are 

small. Both the maximum and average annualized concentrations are used to estimate intakes and 

other assumptions are biased to be favorable to the claimant. An example is the dilution factor used 

to adjust stack effluent concentrations for use as ground-level exposures. Actual data would suggest 

that roof-level stacks tend to create dilution factors of 0.001 or more (ORAUT 2004, p. 28), but in this 

instance a factor of 0.01 is assumed, which results in higher ground-level exposure to an unmonitored 

worker.
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5.0

5.1 INTRODUCTION

OCCUPATIONAL INTERNAL DOSE

BNL has played an important role in the development of the U.S. nuclear program. It has conducted 

applied research in nuclear and high-energy physics, chemistry and physics of materials, 

environmental and energy research, nonproliferation, neurosciences and medical imaging, and 

structural biology. Methods and concepts of measuring occupational internal doses to workers have 

evolved since the beginning of BNL operations. This section describes BNL internal dosimetry 

systems and practices.

5.2 PURPOSE

The purpose of this section is to describe internal dosimetry systems and practices at BNL. This 

section provides supporting technical data to evaluate internal occupational doses that can 

reasonably be associated with worker radiation exposures covered by EEOICPA.

5.3 SCOPE

This section presents historical and current practices as they relate to the evaluation of internal 

exposure data for monitored and unmonitored workers. It describes plant facilities and processes, 

and historical information in relation to dose reconstruction for BNL workers. In addition, it contains 

supporting documentation to assist in the evaluation of occupational internal doses from these 

processes in accordance with the Internal Dose Reconstruction Implementation Guideline (NIOSH 

2002) and addresses the evaluation of dose for unmonitored and monitored workers. 

The section presents the technical basis of methods used to prepare dose information for input to the 

NIOSH IREP computer code, and evaluates the uncertainty for BNL exposure and dose records.

NIOSH has determined, and the Secretary of Health and Human Services has concurred, that it lacks 

sufficient personnel or area monitoring data, or sufficient source or source term information, 

associated with BNL operations to bound potential internal exposures for the period from January 1, 

1947, through December 31, 1979 (other than tritium after December 31, 1964).

NIOSH found that while it is not possible to completely reconstruct internal radiation doses for 

employees who worked at BNL from January 1, 1947, through December 31, 1979, NIOSH intends to 

use any reliable internal monitoring data that may be available for an individual claim (and that can be 

interpreted using existing NIOSH dose reconstruction processes or procedures) to support a partial 

dose reconstruction for nonpresumptive cancers and/or cases that have less than 250 work days of 

employment.

5.4 BIOASSAY RECORDS IN THE INDIVIDUAL FILES

5.4.1

The BNL bioassay record, Form 1720, contains one line (record) per sample. The record includes the 

following fields:

Bioassay Record (Form 1720)

1. NO. (number, i.e., the sequential number of the sample),

2. DATE RECV’D (date sample was received),

3. I or S (code for incident or survey. Survey is assumed to be a routine sample), 

4. COLLECTION DATE/TIME,

5. COUNT DATE, 
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6. TOTAL SAMPLE (volume in milliliters or X for a spot sample), and

7. SAMPLE TYPE (U for urine, F for feces, B for blood, H for hair, and WB for whole-body).

The results portion of the form is filled in according to the analysis. Columns are provided for MFPs

(mixed fission products in dpm/day beta activity), 90Sr (in dpm/day beta activity), and 3

H (in μCi/L beta 

activity). Additional columns are provided for NUCLIDE, AMOUNT, and UNITS for alpha activity. 

Results for 90Sr appeared only with MFP results, indicating a sequential analysis. This form appears 

to have been used from 1967 to the 1980s. Data from as far back as 1952 appear to have been 

transferred to these forms in 1967.

5.4.2

This form was used to calculate the whole-body dose from tritium bioassay results. Results for the 

bioassay sample under evaluation that were above some threshold (probably around 1 μCi/L) were 

transferred to this form. Each row was used to calculate the dose (if any) from one urine result. The 

column headings are:

Tritium Exposure Evaluation Form

1. No. of Intervening Days (since last sample)

2. Gross μCi/L Current Date

3. Gross μCi/L Previous Date

4. Fraction Present of Previous Concentration

5. μCi/L Remaining from Previous Date

6. Net μCi/L Current Date

7. Tritium Body Dose for Current Date MREM

8. Cumulative Tritium Body Dose MREM

Doses were not calculated if the Net μCi/L Current Date (Gross μCi/L Current Date minus μCi/L 

Remaining from Previous Date) was less than some threshold. The threshold might have been 0.3 

μCi/L. The form was used from the fourth quarter of 1968 until at least 1986. The doses were totaled 

for each quarter for comparison to the limits in place at the time. This form could be a secondary 

source of information should the BNL Bioassay Record (Form 1720) discussed above be missing or 

illegible. Based on the review of site records and personnel interviews, tritium dose was recorded on 

the external dosimetry cards (in the employee records) until the committed dose era starting in 1989.

5.4.3

Summaries of the whole-body counting results might be included in some employees’ records for the 

1970s and possibly earlier. The summaries currently list several employees on a single page. Cover 

letters included with the results provided an interpretation of positive results. For example, the letters 

discuss cesium results believed to be from fallout, results attributed to clothing contamination, results 

due to an intake with a body burden calculated, etc. [12].

Summary of Whole-Body Counting Results

5.4.4

By the 1980s, whole-body counting results for each count were included in the records. The earliest 

results were titled “NAI Results” and consisted of a one- or two-page computer printout. The counts 

appeared to be done with a single detector and were 15 minutes long. The printout contains:

Individual Whole-Body Counting Results

1. ACQUISITION DATE, date and time of the count 

2. ELAPSED LIVE TIME, length of the count 

3. NUCLIDE, a table of results listing the radionuclide 
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4. CONCENTRATION (UNITS), activity 

5. ERROR, error

By 1993, the site had changed to the Canberra ABACOS-plus software package. The ABACOS 

printout contained:

1. COUNT STARTED, the date and time of the count 

2. ELAPSED LIVE TIME, length of the count 

3. NUCLIDE, a table of results listing the radionuclide 

4. ACTIVITY (nCi), the activity 

5. %ERROR (2 SD), and the error 

Initially the ABACOS printouts displayed “<” and the minimum detectable activity (MDA) in the results 

column for results below the MDA. Due to software limitations, an MDA and decision level could not 

both be displayed. In preparation for DOE Laboratory Accreditation Program (DOELAP)

accreditation, in 1999 the number displayed was changed to the decision level, but the printout 

incorrectly stated that the number displayed was the MDA (i.e., “MDA activity reported”). A list of 

MDAs was published separately. This was corrected when the software was upgraded and 

subsequently both a decision level and an MDA were reported [13].

5.5 IN VITRO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PROTOCOLS

5.5.1

In January 1949, a chemistry laboratory was organized under the Health and Safety Branch, Medical 

Division, New York Operations Office (NYOO) of the AEC to provide analytical and spectrographic 

facilities for dealing with hazards arising from NYOO facilities (e.g., uranium, beryllium, fluorine, 

thorium, and radium). In April 1949, the NYOO chemistry laboratory began an investigation into the

feasibility of using the levels of uranium excretion in urine as a control measure in production plants 

and processing facilities (AEC 1949). In 1953, this organization became the Health and Safety 

Laboratory (HASL) and in 1977 the Environmental Measurements Laboratory (EML). In 2002, it 

became part of the Department of Homeland Security. 

In Vitro Urine Analysis

5.5.2

Initially, support was also obtained from Oak Ridge National Laboratory to perform urinalysis for 

plutonium, polonium, mixed fission products, barium-lanthanum, yttrium, and strontium. By 1951, 

BNL was doing its own research and development on urine sample preparation (BNL 1951). In 1952, 

an analysis procedure for fission products in urine was successful using an adaptation of the Chalk 

River method (BNL 1952). However, in 1955, some uranium in urine at BNL was still analyzed 

fluorometrically by EML (AEC 1955). In 1958, a vibrating reed electrometer and chambers suitable for 

tritium urinalysis were set up in Building T-145 (Bishop 1958). In 1965, enriched uranium in urine was 

analyzed by EML. The analysis appears to have used both liquid scintillation counting (perhaps as a 

screening measurement) and electroplating for counting or autoradiography (Steimers 1965). The 

uranium fusion photofluorimetry urinalyses performed by the NYOO were similar to those performed 

at other AEC facilities. The default detection threshold for uranium urinalysis is assumed to be 10 

μg/L based on a reported sensitivity of 5 to 10 μg/L for uranium fluorometric urinalysis in the early 

History of In Vitro Urine Analysis
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 69 of 115

years (ORAUT 2005). During the period from 1956 to 1964, the reporting level for uranium in urine by 

fluorometric analysis by HASL appears to have been 1.0 μg/L (see for example AEC 1957). 

Determination of 233U and 235U in urine was by fluorometry and alpha counting (AEC 1958, 1961, 

1962). It is uncertain whether the analyses were for direct support, verification, or only for specific 

analyses needed. In May 1963, 112 urine samples were processed by the in-house bioassay 

laboratory. Gross beta and 90Sr activity was measured in one individual who had been contaminated 

in April during reloading of spent fuel elements for shipment (BNL 1963).

In 1973, bioassay analyses were performed by an Analytical Chemistry section (BNL 1973). In 1977, 

the Health Physics and Safety Division housed an analytical chemistry and bioassay laboratory 

(ERDA 1977). In 1999, Eberline was contacted for bioassay analyses, with the exception of 3

H, which 

remained in-house until 2004 (~April 2004). Due to some problems with meeting the contractual 

MDAs, Eberline was replaced by the Severn-Trent (ST) Laboratory in December 2001. ST also did 

the H-3 analyses when they were discontinued on site. The current BNL bioassay program is 

accredited by DOELAP.

During the operation of the BGRR, some workers apparently were routinely sampled for mixed fission 

products and 90Sr. No program documentation was recovered but this seems to be the case from a 

limited review of the individual files. Sampling frequency varied from as short as monthly to as long 

as annually. After 1965, when the HFBR started operation, the most prevalent bioassay sampling 

was for tritium. During the operation of the HFBR, the concentrations of radionuclides other than 

tritium were not considered to be significant. The rationale was that concentration of tritium in the 

primary cooling water was several million times its “tolerance limit” where fission products from fuel 

uranium surface contamination and activations products, such as 24Na, were present at less than 100 

times tolerance (BNL undated, p. 14). 

During most of the site’s existence, it appears that department/division safety representatives or 

managers designated personnel for participation in the internal dosimetry program [whole-body 

counters (WBCs) and/or urinalysis] using guidance provided by Personnel Monitoring. This could 

have included the frequency of monitoring (BNL 1995; Holeman 1999). Therefore, it might be difficult 

to determine if monitoring should have occurred by job title. For example, in the 1950s, several 

references mentioned that personnel described as “janitors” were involved in the cleanup of 

radioactive contamination. It was not clear if this was a specially designated crew or random site 

janitors being supervised by health physics personnel.

As some of the site’s older facilities have undergone decontamination and decommissioning (D&D) in 

recent years, D&D workers were sampled for 90Sr, 241Am, 238Pu, 239/240Pu, 234U, 235U, and 238U.

5.5.3

Although fecal sampling might have been used in response to specific events, no evidence of a 

comprehensive fecal sampling program was located.

Fecal Sample Analysis

Table 5-1 lists the frequencies for in vitro monitoring.

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

Routine monitoring type Period Frequency

Spot urine sample for H-3 Before 

1993

No comprehensive documentation was located. It appeared to 

vary from weekly to monthly over time. The majority of samples 

after 1965 were from workers at the HFBR

Urine samples for H-3a 1993–

2000

Weekly for reactor operators, supervisors, maintenance 

technicians, and S&EP building safety services technicians
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 70 of 115

24-hr urine samples for fission, 

activation, uranium, and 

transuranics

No comprehensive documentation was located. In 1984, an 

annual program of urine sampling was in place for reactor 

workers. The analysis included Cd-109, Sr-85, Ga-68, Na-22, 

and Co-58b.

a. BNL (1993).

b. Miltenberger and Steimers (1984).

5.6 IN VIVO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PRACTICES

5.6.1

Whole-body counting was established in 1960. It was first used as a research tool, but it became 

evident that it was also of use when someone was potentially involved in an incident. After confirmed 

intakes, WB counting was repeated daily and the results were correlated with daily urine excretion 

results. Pile operators were the first group to be routinely counted. Some of the pile operators had 

been working before WB counting started. Later, all new pile operators received a WB count before 

they started working and, starting in about 1962, annually thereafter. Other groups were apparently 

counted in the WBC as time permitted. Background counts were taken on unexposed individuals as a 

check on fallout radionuclides (e.g., cesium) in the general population (Brodsky 1964, p. 40). The 

counting facility, which is still in existence in the Medical Department, includes a counting bed with an 

array of sodium iodide thallium-activated [NaI(Tl)] scintillation crystals.

Whole-Body Counting

By 1963, a portable “shadow shield” WBC was developed to measure the fission product body 

burdens of the Marshallese people who were exposed to the fallout from nuclear test Bravo on March 

1, 1954, and to the subsequent uptake of radioactive material in crops. The detector was a 10-cmthick by 28-cm-diameter NaI(Tl) crystal. The detector was stationary and was fixed in a position over 

the thorax of the subject. For more than 25 years, BNL had a contract for the long-term medical 

surveillance of the Marshallese. In 1977, the responsibility for providing body burden measurements 

was transferred from the Medical Department to the Safety and Environmental Protection Division. In 

1980, the minimum detection limit of the system was 1 nCi of 137Cs or 60Co for a 15-minute count 

(Miltenberger, Greenhouse, and Lessard 1980). Between these measurement trips, the counter was 

used at BNL as part of the health and safety program and eventually performed the routine counts 

(Miltenberger 1981a). For positive results, additional counts could be done at the medical WB

counting facility to localize the contamination (Miltenberger 1981a). The MDL for the shadow shield 

arrangement was calculated in μCi by:

K

Ct

MDL b

900 * 37 *

4.65 = (5-1)

where:

Ctb = blank counts (background for region of interest)

900 = count time (seconds)

37 = unit conversion factor

K = product of gamma abundance and counting efficiency for the radionuclide

By 1981, 137Cs body burdens of 2 nCi or less were considered to be the result of fallout from 

atmospheric weapons testing environmental sources (Miltenberger 1981b).

In 1989, the MDA reported for thyroid counts for 131I using a NaI counting system was “on the order of 

10 nanocuries” (Miltenberger 1989).
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 71 of 115

The shadow shield was replaced by a standup counter, but the date of this transition was not 

recovered. The NaI detectors were replaced by germanium detectors and eventually by broad energy 

germanium detectors.

In 2003, Canberra ABACOS 2000 software was used with the WBC. The MDA for each radionuclide 

was calculated by:

KT

S MDA 3.29 * b + 3 = (5-2)

where:

Sb = sample standard deviation of background counts in the region of interest

K = efficiency at the centroid energy channel as determined arithmetically from the 

calibration efficiency equation times the yield abundance of the energy emission

T = count time

The ABACOS 2000 software produced a printout with the MDA and decision level for each result 

(Michel and Sun 2003; Michel 2003, 2004, 2005).

5.6.2

Chest counts and thyroid counts were not done routinely. These counts were done as a result of 

incidents involving iodine (thyroid) and certain gamma emitters (127Xe, lung) (Miltenberger and 

Lessard 1987). Only one person who does iodinations was on a thyroid counting program in 2006.

Chest Counting

Table 5-2 lists the frequencies for in vivo monitoring.

Table 5-2. Internal dose control program (in vivo).

Routine monitoring type Period Frequency

Preemployment and annual 

WBCa 1962–unknown Annual for pile operators. Other workers designated by the facility health physics staff of the various facilities.

WBCb 1993–2000 Start of service in Bldg. 750, and termination of service for 

Reactor Division and S&EP personnel with offices in the 

HFBR building (750). Annual for reactor operators, 

supervisors, maintenance technicians, S&EP building safety 

services technicians and research coordination group 

technicians, BLIP personnel, facility support, others on prejob 

basis.

a. Brodsky (1964)

b. BNL (1993)

5.7 UNCERTAINTY

At BNL, the uncertainty for a single bioassay measurement was not reported consistently. 

Uncertainties associated with the analysis of MFPs, 90Sr, or 3

H performed on site during the period 

from 1952 to 1986 were not included on the BNL Bioassay Record. From the 1980s to about 1998,

uncertainties are reported on WBC sheets and are believed to be consistently reported as 2-sigma 

errors. Reviewed statements of work for bioassay services do not contain any specification for 

reporting uncertainty. Eberline reported a 1-sigma total error with its bioassay results. At present, the 

ST contract laboratory reports results with both a 1-sigma counting error and a 1-sigma total error 

[14].
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 72 of 115

5.8 DETECTION LIMITS

Table 5-3 lists the MDAs and reporting levels for periods corresponding to the bioassay methods 

discussed in Sections 5.5 and 5.6. The reporting levels are listed in the units quoted in the 

references, which are generally the units of the results. Detection limits are to be taken from the 

individual reports when available.

5.9 EXCRETA SAMPLE KIT CODES

No codes have been found. Table 5-4 lists sample kit information summarized from site documents.

Table 5-3. Detection limits.

Radionuclide

Method/

description Period MDA (nCi)a

Reporting level

(nCi)b

Be-7 (dual detector) WBC 1993 50 50

Mn-54 (dual detector) WBC 1993 4.7 4.7

Cs-137 (dual detector) WBC 1993 5 5

Co-60 (dual detector) WBC 1993 5 5

Zn-65 (dual detector) WBC 1993 8.5 8.5

I-131 (dual detector) WBC 1993 3.8 3.8

Be-7 (single detector) WBC 1993 82 82

Mn-54 (single detector) WBC 1993 7.5 7.5

Cs-137 (single detector) WBC 1993 8.5 8.5

Co-60 (single detector) WBC 1993 9.2 9.2

Zn-65 (single detector) WBC 1993 17.6 17.6

I-131 (single detector) WBC 1993 8.8 8.8

a. Values from Murray (1993).

b. Reporting level was the MDA.

Radionuclide (keV)

Method/

description Period MDA (nCi)a Reporting levelb

Na-22 (1,274.54) WBC 1999 3

Mn-54 (834.83) WBC 1999 3

Co-57 (122.06) WBC 1999 7

Co-58 (810.76) WBC 1999 4

Fe-59 (1,099.22) WBC 1999 4

Co-60 (1,332.49) WBC 1999 5

Zn-65 (1,115.52) WBC 1999 5

Cs-134 (604.7) WBC 1999 3

Cs-137 (661.65) WBC 1999 4

a. Values from Murray (1999).

b. Reporting levels shown on lung count results reports are from ABACOS-Plus software.

Radionuclide (keV)

Method/

description Period MDA (nCi)a Reporting level

Be-7 (477.59) WBC 2003 21.96, 22.2

Na-22 (1274.54) WBC 2003 2.58, 3.21

K-40 (1,460.81) WBC 2003 36.50, 131.54

Mn-54 (834.83) WBC 2003 2.22, 3.74

Co-57 (122.06) WBC 2003 3.37, 2.44

Co-58 (810.76) WBC 2003 2.49, 3.23

Fe-59 (1,099.22) WBC 2003 4.04, 6.07

Co-60 (1,332.49) WBC 2003 1.83, 3.65

Zn-65 (1,115.52) WBC 2003 5.18, 6.45

Sr-85 (513.99) WBC 2003 2.38, 5.14

I-129 (29.62) WBC 2003 7.97, 3.01
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 73 of 115

Radionuclide (keV)

Method/

description Period MDA (nCi)a Reporting level

I-131 (364.48) WBC 2003 2.51, 3.47

Cs-134 (604.7) WBC 2003 2.09, 3.08

Cs-137 (661.65) WBC 2003 2.64, 4.77

Ce-139 (165.85) WBC 2003 3.48, 2.93

Ce-144 (133.54) WBC 2003 27.12, 22.11

Bi-214 (609.31) WBC 2003 7.00, 9.65

Pb-214 (351.92) WBC 2003 7.02, 8.5

Am-241 (59.54) WBC 2003 9.06, 4.94

a. Values for 2003 from Michel (2003) and Michel and Sun (2003); MDAs are for each calibration.

Radionuclide (keV)

Method/

description Period MDA (nCi)a Reporting level

Be-7 (477.59) WBC 2004 23.35

Na-22 (1,274.54) WBC 2004 3.97

K-40 (1460.81) WBC 2004 147.04

Mn-54 (834.83) WBC 2004 2.50

Co-57 (122.06) WBC 2004 1.40

Co-58 (810.76) WBC 2004 2.39

Fe-59 (1099.22) WBC 2004 6.10

Co-60 (1332.49) WBC 2004 4.12

Zn-65 (1115.52) WBC 2004 5.09

Sr-85 (513.99) WBC 2004 2.21

I-129 (29.62) WBC 2004 78.83

I-131 (364.48) WBC 2004 2.20

Cs-134 (604.7) WBC 2004 2.29

Cs-137 (661.65) WBC 2004 3.62

Ce-139 (165.85) WBC 2004 1.94

Ce-144 (133.54) WBC 2004 13.59

Bi-214 (609.31) WBC 2004 11.69

Pb-214 (351.92) WBC 2004 8.32

Am-241 (59.54) WBC 2004 5.73

a. Values for 2004 from Michel (2004) published for each calibration.

Radionuclide (keV)

Method/

description Period MDA (nCi)a Reporting level

Be-7 (477.59) WBC 2005 20.13

Na-22 (1274.54) WBC 2005 2.76

K-40 (1460.81) WBC 2005 42.99

Mn-54 (834.83) WBC 2005 2.79

Co-57 (122.06) WBC 2005 1.91

Co-58 (810.76) WBC 2005 2.91

Fe-59 (1099.22) WBC 2005 4.92

Co-60 (1332.49) WBC 2005 2.90

Zn-65 (1115.52) WBC 2005 5.54

Sr-85 (513.99) WBC 2005 3.41

I-129 (29.62) WBC 2005 81.99

I-131 (364.48) WBC 2005 1.92

Cs-134 (604.7) WBC 2005 2.89

Cs-137 (661.65) WBC 2005 3.12

Ce-139 (165.85) WBC 2005 1.76

Ce-144 (133.54) WBC 2005 14.63

Bi-214 (609.31) WBC 2005 7.72
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 74 of 115

Radionuclide (keV)

Method/

description Period MDA (nCi)a Reporting level

Pb-214 (351.92) WBC 2005 7.60

Am-241 (59.54) WBC 2005 4.42

a. Values for 2005 from Michel (2005) published for each calibration.

Radionuclide

Method/

description Period MDAa Reporting levelb

Am-241 (Eberline) Urinalysis 1999–2000 6.0 × 10-2 pCi/L No threshold

Am-241 Urinalysis 2001–2006 8.4 × 10-8 μCi/24 hr No threshold

Br-77 Urinalysis 1975 5 μCi/Lc

H-3 (in house lab)d Urinalysis 1965–1979 2.0 × 10-2 μCi/L

H-3 (in house lab) Urinalysis 1996 4.0 × 10-3 μCi/L

H-3 (in house lab) Urinalysis 1999 3.3 × 10-3 μCi/L

H-3 (in house lab)e Urinalysis 2000 5.0 × 10-3 μCi/L

H-3 (required by TBD) Urinalysis 2001–2006 1.0 × 10-2 μCi/L

Mixed fission productsd Urinalysis 1952–1969 5.0 dpm/24 hr

Pu-238 (Eberline) Urinalysis 1999–2000 6.0 × 10-2 pCi/L No threshold

Pu-238 Urinalysis 2004–2006 8.4 × 10-8 μCi/24 hr No threshold

Pu-239 Urinalysis 2001–2006 8.4 × 10-8 μCi/24 hr No threshold

Pu-239/240 (Eberline) Urinalysis 1999–2000 6.0 × 10-2 pCi/L No threshold

Sr-90c Urinalysis 1950–1998 5.0 dpm/24 hr

Sr-90 (Eberline) Urinalysis 1999–2000 1.0 pCi/L No threshold

Sr-90 (required by TBD) Urinalysis 2001–2006 3.54 × 10-2 μCi/24 hr No threshold

Uranium (fluorometric) Urinalysis Up to 1955 10 μg/L

Uranium (fluorometric) Urinalysis 1956–1965 1 μg/L

U-238 Urinalysis 2001–2006 8.4 × 10-8 μCi/24 hr No threshold

U-238 (Eberline) Urinalysis 1999–2000 1.0 × 10-1 pCi/L No threshold

U-235 Urinalysis 2001–2006 8.4 × 10-8 μCi/24 hr No threshold

U-235 (Eberline) Urinalysis 1999–2000 1.0 × 10-1 pCi/L No threshold

U-234 (Eberline) Urinalysis 1999–2000 1.0 × 10-1 pCi/L No threshold

a. Values from BNL (2001b) and Sun, Reciniello, and Sengupta (2004), except as noted. Values could be applicable 

earlier, but documentation was not located.

b. Reporting levels shown were not documented as the MDA but can be interpreted as the MDA in lieu of other 

information.

c. Value below which “no exposure was assigned” from O’Connell (1975).

d. Values from a reactor employee’s bioassay data.

e. Value from Scarpitta (2001).

Table 5-4. Excreta sample kit codes.

Kit codea

D/R P/U Media Sample description

NA NA Urine Routine samples for H-3 were collected as grab samples (1993) to be collected 

weekly or within 4 hr of completion of a task that required bioassay. Twenty-fourhour samples could also be collected when other radionuclides were suspected.a

 In 

1984, 24-hour kits consisted of three 500-mL bottles.b

 In 1999, spot samples were 

collected in 60-cc plastic bottles.c

a. BNL (1993).

b. Miltenberger and Steimers (1984).

c. Holeman (1999).

5.10 SOLUBILITY TYPE, FRACTION ACTIVITY, AND PARTICLE SIZE BY FACILITY

In the absence of measurements or studies, NIOSH guidance requires the use of default solubility 

classes and particle size values from the ICRP (NIOSH 2002, pp. 15, 16). Facility-specific solubility 

and particle size data for BNL have not been found. Activity fractions for occupational exposure were 
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 75 of 115

generally not available. For some facilities, stack emissions or other measurements could be useful 

for dose reconstructions. Table 5-5 lists this information.

Table 5-5. Solubility type, fraction activity, and particle size by facility.

Facilitya Compoundb RN

Solubilityc

type

Particle sized

(μm AMAD)

Activity 

fraction

Sewage 

Treatment Plant

(1962)

Unknown. Measured at the outfall 

from the Imhoff tank (Stubbings 

1962)

Sr-90 0.20

Cs-137 0.80

Sewage 

Treatment Plant

(1973)

Unknown. Fractions based on 

activities measured at the input to 

the sand filter beds (ERDA 1977, p. 

126).

H-3 9.87E-01

Be-7 7.33E-03

Na-22 6.01E-05

Na-24 3.59E-03

Cr-51 2.49E-04

Co-58 9.02E-05

Co-60 2.58E-05

Zn-65 3.44E-05

Sr-90 5.58E-05

I-131 1.37E-03

Cs-134 1.72E-05

Cs-137 1.07E-04

Ce-144 1.98E-04

BGRR

Bldgs. 701-703, 

704 (fanhouse), 

708 (instrument 

house), 709 

(canal house), 

709A (canal 

water treatment 

facility) (1950–

1957)

Natural uranium metal fuel, 1-in, × 

4–in. cylinders in 11-ft finned 

aluminum cartridges.

(Activity fractions based on 

estimates of core activity at the 

time of 28 fuel ruptures and one 

irradiated uranium sample failure) 

(Meinhold and Meinhold, 2001)

Ba-140 1.70E-01

Ce-144 1.08E-01

Cs-137 7.19E-03

La-140 1.95E-01

Nb-95 1.62E-01

Ru-103 7.28E-02

Ru-106 6.09E-03

Sr-89 1.17E-01

Sr-90 7.78E-03

Zr-95 1.53E-01

U-235 8.35E-08

U-238 1.81E-06

U-234 1.81E-06

Pu-239 2.86E-04

BGRR

Bldgs. 701-703 

(1958–1969)

Enriched uranium fuel, in the form 

of curved plates of uraniumaluminum alloy, clad on all surfaces 

by 0.5 mm of Al. Routine releases 

of fission products occurred due to 

U contamination on the fuel’s 

surface and trapped in the 

cladding. Based on I-131 and 

gross beta measurements in the 

1967–1969 EMRs.

Co-60 8.05E-02

Zr-95 3.79E-01

Ru-103 2.30E-01

Ce-141 1.26E-01

Ce-144 1.15E-01

Cs-137 6.90E-02

I-131 2.86E-03

BGRR

Bldgs. 701-703 

(1970–2006)

Residual contamination resulting 

from fission, activation, and 

transuranic radionuclides produced 

in the core and spread primarily by 

contact with the cooling water 

(Musolino 2000). Fuel storage 

canal was decommissioned in 

1973.

Pu-238 6.66E-06

Pu-239/40 5.86E-04

Pu-241 2.59E-04

Am-241 5.05E-05

Fe-55 9.79E-01

Co-60 4.89E-04

Sr-90 9.09E-03

Cs-137 1.08E-02

U-235 1.37E-07

U-238 3.57E-06
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 76 of 115

Facilitya Compoundb RN

Solubilityc

type

Particle sized

(μm AMAD)

Activity 

fraction

BMRR

Bldgs. 490-491 

(1959–2000)

Enriched uranium fuel, in the form 

of curved plates of uraniumaluminum alloy, containing 12% by 

weight of fully enriched uranium. 

Estimates are for 1997 routine 

releases of isotopes likely to be of 

dosimetric importance (Meinhold 

and Meinhold 2001, Table 5, p. 57) 

99.9% of activity release to sewer 

from labs was H-3 (ERDA 1977, p. 

123)

Ce-141 2.99E-03

Ce-144 2.33E-02

Co-60 4.35E-02

Fe-59 6.32E-02

I-131 5.47E-01

Zn-65 3.20E-01

Chemistry

Hot Chemistry

Polonium Facility

T-137

1954–1961

Polonium source used in chemistry 

apparatus; also separated as a byproduct of the production of At-211 

for medical purposes

Po-210

Cosmotron Bldg. 

(1953–1966)

Activation of short-lived isotopes in 

the ventilating air in the tunnel and 

experimental areas. See 

”Accelerators & RHIC” below.

C-11

N-13

O-15

AGS

Bldgs. 905-912 

(1960–2006)

Activation of short-lived isotopes in 

the ventilating air in the tunnel and 

experimental areas. Processing of 

irradiated targets (Meinhold and 

Meinhold, 2001, p. 58; ERDA 1977, 

p. 310). See ”Accelerators & 

RHIC” below.

C-11

N-13

O-15

AGS Bubble 

Chamber 

Bldg. 919

Handling of tritium sources up to 

100 mCi (1977)

H-3 1

Accelerators and 

RHIC

1953–2006

Posted contamination areas 

primarily in the vicinity of fixed 

targets (Lessard, Schaefer, and 

Karol 2001). Based on recent 

information, but is assumed to 

apply to earlier routine operations. 

Before 1987, beam intensities were 

less but some uncontained targets 

(e.g., tungsten) were used.

Be-7 5.50E-01

Na-22 1.40E-01

Mn-54 1.40E-01

Co-57 1.40E-01

Co-60 3.00E-02

Hot Laboratory

(central lab, 

fanhouse, liquid 

waste tank farm, 

liquid waste 

concentration 

plant)

Bldg. 801, 802

(1951–2006)

Acid dissolution of irradiated 

uranium samples to recover 

radioiodines (March 1952 to June 

1960)

In 1957, there was an accidental 

release of unirradiated natural 

uranium hexafluoride due to an 

explosion (Meinhold and Meinhold, 

2001, p. 59).

Radioactive solutions up to mCi 

amounts, 200 Ci/yr of H-3 

fabricated into accelerator targets 

(1977). 87% of the liquid waste in 

1973 was H-3 (ERDA 1977, p. 125)

Fractions based on environmental 

releases of isotopes of dosimetric 

importance, 1983–1991

Co-58 9.33E-02

Co-60 1.30E-03

Cs-137 1.92E-02

Eu-155 3.37E-03

I-131 8.67E-02

Mn-54 1.88E-03

Zn-65 7.91E-01

Zr-95 2.95E-03

U Unknown

H-3 Unknown
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 77 of 115

Facilitya Compoundb RN

Solubilityc

type

Particle sized

(μm AMAD)

Activity 

fraction

Hot Laboratory

(1955–1966)

Polonium contamination; Po-210 

sources manufactured in alpha 

facility

Po-210

Nuclear 

Engineering

T-197, 480

(1954–1962)

Unspecified polonium use requiring 

urinalysis (1954); and air sampling 

(1962)

Po-210

Physics

T-248, T-109, 

others before 

(1947–1951)

Polonium used in physics 

experiments

Po-210

Physics

Bldg. 510

(1962–present)

Processing irradiated targets, up to 

1 mCi; 5,000 Ci of Co-60 in the 

gamma irradiation facility (1977)

Radiation and 

Chemical

Technology

Bldgs. 526-527

Unknown uranium of various 

enrichments. Originally housed a 

criticality facility for reactor physics 

(ERDA 1977, p. 48). Monomer 

processing (ERDA 1977). 

Concrete dried and impregnated 

with monomers.

HIRDL

Bldg. 528

(1963–1972)

Contained million-Ci Co-60 and Cs137 sources. The HIRDL was used 

for the development and testing of 

these sources. Operations were 

curtailed in 1972 due to lack of 

funding. In 1977, only a water-tank 

Co-60 gamma irradiation facility 

was still in operation.

Cs-137

Co-60

Radiation 

Development 

Laboratory

Bldg. 830

Fabrication of solid sources, up to 

Ci amounts, monomer processing 

(1977)

Hot Machine 

Shop

Bldg. 530 

–1975

Replaced by

Bldg. 462

Machining radioactive solids 

(1977), including uranium and 

radioactive graphite; repair work on 

“hot” fuel- handling tools.

Health Physics & 

Safety, 

Instrumentation

Bldg. 535

Radioactive solutions up to 100 μCi 

(1977)

Chemistry

Bldg. 555

Radioactive solutions and irradiated 

targets, up to 1 mCi (1977)

CLIF

Bldg. 931A

Production of radioactive isotopes. 

See ”Accelerators & RHIC” above. 

Dates of operation were not 

located.

Reclamation and 

Hot Laundry

Bldg. 650

Radioactive decontamination, up to 

mCi amounts (1977)

82% of activity released to sewer in 

1973 was H-3, rest was gross beta 

(ERDA 1977, p. 125)
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Facilitya Compoundb RN

Solubilityc

type

Particle sized

(μm AMAD)

Activity 

fraction

HFBR

Bldg. 750

(1965–1997)

“Fully” enriched uranium and 

aluminum alloy, heavy-watercooled and -moderated reactor that 

operated up to 40 MW (thermal) 

(60 MW planned in 1977)

Fractions based on activity in heavy 

water when systems were opened 

for refueling or maintenance 

(1977).

H-3 9.998E-01

Na-24 2.22E-04

Cr-51 2.22E-06

Co-59 8.89E-10

Co-60 6.67E-10

Zn-69 8.89E-09

I-131 6.67E-10

Cs-137 3.33E-10

Tandem 

Van de Graaff 

Accelerator

Bldg. 901, 901A

Tritium gas and vapor. Fractions 

based on activity released from 

facility in 1979 (Naidu 1979)

H-3 Gas 8.95E-01

H-3 Vapor 1.05E-01

Superconducting 

Test, Machine 

Shop, RARAF

Bldg. 902

Irradiated target processing, up to 

100 μCi. (1977)

BLIP Facility

Bldgs. 930 and 

931B

Processing irradiated samples 

containing short-lived radionuclides 

(1977) for medical use; targets 

were sealed during normal 

operations but radioactive gases 

(N-13, O-14, O-15, N-16) and other 

isotopes formed in cooling water 

and equipment (ERDA 1977, p. 

113). Fractions estimated based 

on activity in cooling and shield 

tank after one year of operation 

(ERDA 1977, p. 123).

H-3 HTO 6.35E-01

Be-7 3.05E-01

Na-22 2.54E-03

Na-24 5.08E-02

Cr-51 4.07E-03

Co-58 1.52E-03

Co-60 1.02E-03

Waste 

Management 

Incinerator

Bldg. 445

(1981–)

BNL granted a permit by New York 

in March 1981. Fractions based on 

averages 1981–1991 for isotopes 

of importance for internal dose.

Fe-59 9.01E-02

I-131 1.03E-01

Mn-54 1.02E-02

Ru-103 2.87E-02

Ru-106 2.68E-03

Tc-99 5.01E-02

Zn-65 7.15E-01

a. Facilities were combined for this analysis if they were similar and had a common list of radionuclides of concern.

b. If chemical compounds were not available, the best description found is listed.

c. Specific information was not located; the solubility that is most favorable to the claimant should be applied.

d. Specific information was not located; the 5-μm default is assumed to apply.

5.11 FACILITY-SPECIFIC RADIONUCLIDE CONVERSIONS

The natural uranium fuel elements for the BGRR were reportedly fabricated by the BNL metallurgy 

group. The enriched uranium, aluminum-clad elements that replaced the natural fuel were apparently 

not fabricated at BNL. These data are summarized in Table 5-6.

Table 5-6. Facility-specific radionuclide conversions.

Process description

Activity per unit mass (Bq/g uranium)

U-234 U-235 U-236 U-238

BNL metallurgy group, natural U fuel fabrication, late 

1940s and possibly up to 1958

1.26E+04 5.59E+02 negligible 1.21E+04
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5.12 WORKPLACE MONITORING DATA

If bioassay data are not adequate to evaluate an individual’s internal doses, dose reconstructors can 

use workplace monitoring data (NIOSH 2002). The following types of workplace data might be 

available for BNL: breathing zone air samples, general area air samples, and surface contamination 

surveys. However, these data are not likely to be in individual exposure records. Data on respirator 

use are not likely to be available. Fit testing records are Industrial Hygiene records maintained by the 

Safety and Health Services Division. In the case of surface contamination data, site/process-specific 

resuspension factors are not likely to be available. During the operation of the HFBR, the airborne 

tritium contamination had been between 1 × 10-6 μCi/cm3 and 1 × 10-5 μCi/cm3 on the equipment level. 

Since the shutdown, the concentration has been between 1 × 10-7 μCi/cm3 and 5 × 10-7 μCi/cm3 on 

the equipment level and between 1 × 10-8 μCi/cm3 and 5 × 10-8 μCi/cm3 on the operations level. A 

reasonable estimate of the maximum exposure can be made by assuming that an individual spends 4

hours/day on the operations level and 4 hours/day on the equipment level (BNL 1999).

5.13 SOURCE TERM DATA

Without bioassay or air sample data, the last resort is determination of airborne concentrations using 

source term evaluations (NIOSH 2002, p. 19). Data on the amount of dispersible material available 

does not appear to be available for BNL.

5.14 RADON

For dose reconstruction under EEOICPA, occupational radon exposure is exposure to radon 

emanating from sources other than those naturally occurring in the area. Dose reconstructors must 

subtract the natural background level of radon exposure from any measured values when assessing 

occupational exposure (NIOSH 2002, p. 32). BNL was not a processing or storage location for large 

quantities of 226Ra or 222Rn [15]. On-site environmental monitors occasionally detected low levels of 

226Ra between 1987 and 1996 (see Table 4-2). The source was likely the April 26, 1986 accident at 

the Chernobyl Nuclear Power Plant in Ukraine.

5.15 GUIDANCE FOR DOSE RECONSTRUCTORS

During the ER process NIOSH determined that bioassay data after 1979 was available from the site. 

Therefore if the energy employee has no bioassay data, it will be assumed that the energy employee

didn’t need to be monitored for internal exposure and only ambient environmental intakes need to be 

assigned. The following is some additional guidance for tritium dose assessments. If there is no 

bioassay and no tritium doses reported with external doses, then no tritium dose is assigned for any 

given year. If there is no tritium bioassay, but there is a tritium dose with the external report, use the 

larger of 

• the reported dose (lognormal distribution with a GSD of 3) or 

• the missed dose (triangular dist) based on the threshold at which BNL recorded bioassay 

results or calculated H-3 dose. Based on a review of the records, the threshold for calculating 

dose was 0.3 μCi/L. The annual missed dose based on the application of the 0.3 uCi would be 

11 mrem (0, 0.011 and 0.022 would be the IREP input parameters).
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 80 of 115

6.0

6.1 INTRODUCTION

OCCUPATIONAL EXTERNAL DOSIMETRY

BNL was established to provide facilities for scientific research which, because of size, complexity or 

mode of operation were beyond the means of most single universities. In providing nuclear reactors 

and particle accelerators for its own staff and perhaps more importantly visiting scientists, emphasis 

was placed on a multidisciplinary approach to scientific questions. That same mode of operation is 

still in practice today.

6.2 PURPOSE

The purpose of this section is to describe the external dosimetry systems and practices at BNL, and to 

assist in the evaluation of occupational external exposures from processes that occur at BNL. An 

objective of this section is to provide supporting technical data to evaluate, with assumptions 

favorable to the claimant, occupational external doses that can reasonably be associated with 

radiation exposures to both the monitored and unmonitored worker. 

6.3 SCOPE

This section presents historical and current practices in relation to the evaluation of external exposure 

data for monitored and unmonitored workers, and addresses the evaluation of worker exposure, 

missed dose, and the bias and uncertainty associated with the monitoring of external dose.

6.4 BNL DOSIMETRY PROGRAM

The Oak Ridge dosimeter was used and processed by BNL from start-up through 1984, at which time 

they switched to a vendor (Lane and Reciniello 2003). The vendor services continued through 1995. 

It should be noted that the dosimetry data through 1984 is available only as quarterly summaries, 

even though the badge exchange period was monthly.

In December 1995, BNL was DOELAP-accredited (Loesch 1995) and started its own program using 

the Harshaw 8814 and 8806 TLD dosimeters. At times, primarily in the accelerator facilities, these 

dosimeters were complemented with CR-39 as the neutron dosimeter (Sengupta 2000). 

6.4.1

The neutron doses were measured with NTA film from the beginning of operations through 1995, 

when it was replaced with TLD in combination with CR-39 and Lexan foils. The CR-39 and Lexan are 

analyzed by an outside vendor. Spallation products produced stars on the NTA film, and were 

evaluated by counting the number of prongs on the stars that were produced by those spallation 

products. The total number of prongs were counted and related to an effective neutron dose with the 

result added to the total neutron dose. It was found that the dose from these spallation products was 

very small. A relative biological effectiveness (RBE) and later a quality factor (QF) of 10 was used 

from the very beginning of operations of the accelerators and cyclotrons for these products (Cowan 

ca. 1953).

Neutron Dosimetry

6.4.2

On occasion, special dosimeters were worn to monitor nonroutine work that resulted in “significantly 

non-uniform doses to various areas of the whole body” (Sengupta 2000). The special dosimeters 

were worn without the regular dosimeter. The highest dose measured by the special dosimeters was 

the dose of record. All radiations measured were recorded as WB dose or dose equivalent. The 

Special Dosimetry
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correct value for the RBE (later the QF) was determined to be conservative; because a value of 10 

was always used (Patterson and Thomas 1964; Cowan 1963).

6.4.3

In the cases of lost or destroyed dosimeters, results were derived from past results of similar work, 

coworker results, or the product of instrument measurements and time spent in the radiation zone 

(Reciniello 2006).

Lost or Destroyed Dosimeters

6.5 MISSED DOSE

Missed doses for monitored employees at BNL are basically from dosimeter MDLs and exchange 

periods. In reviewing individual dose records, the exchange period and zeros for that individual can 

be determined. The review of individual records is necessary because exchange periods varied 

before 1955 (Cowan ca. 1953). This is applicable for both Hp (0.07) and Hp (10). 

Another potential source of missed dose historically is possible error in determining true doses from 

high-energy particles. However, it has been determined that the values of the RBEs (and later QFs) 

used historically were conservative resulting in larger than actual true doses being recorded (Cowan 

ca. 1953, Cowan undated). Apparently all workers were issued dosimeters from startup through 1954 

(Cowan ca. 1953).

Table 6-1 lists the period of use, type of dosimeter, exchange period, MDL, and estimated annual 

missed dose.

Table 6-1. Estimated annual missed photon, beta, and neutron dose (mrem). 

Period of use Dosimeter MDLa Exchangefrequency

Maximum annual

missed doseb

Startup–1954 Two-element film 30 Weekly

Monthly

780

180

1955–1995 Multielement film 30 Monthly 180

1996–present Harshaw 8814 TLD 10 Monthly 60

Startup–1995 NTA filmc <50 Weekly

Monthly

1,300

300

1996–present CR-39c <20 Monthly 120

a. Estimated MDLs for each dosimeter in the workplace even though many doses were reported at 

less than the MDL.

b. Estimated annual missed dose calculated using MDL/2 from NIOSH (2007b).

c. Processing done by RS Landauer. Not routinely used after 2003.

6.5.1

Unmonitored workers are those workers who generally did not enter radiation zones or areas where 

work with radioactive materials was undertaken. These workers were not issued dosimeters based 

on the expectation that the probability of exceeding 10% of the allowable limit was small (Sengupta 

2000). Therefore, these workers can be assigned doses as given in Table 4-1 of this document. All 

radiation workers were monitored from the startup of operations (Cowan ca. 1953). 

Unmonitored Worker

Doses for unmonitored employees could be as much as 10% of the relevant standard which was 

1500 mrem/yr (i.e., 300 mrem/wk) in the latter 1940s and then 500 mrem/yr from the early 1950s until 

1987. In 1987, a 100 mrem/yr guideline was established before monitoring was required [16].
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6.6 RADIATION ENERGIES AND PERCENTAGES AT SELECTED BNL FACILITIES

Section 2.0 describes the different types of facilities at BNL, including reactors, accelerators, and 

support facilities and target areas.

Tables 6-2 and 6-3 list the facilities and related data.

Table 6-2. Selection of beta (electron) and photon energies and percentages.

Process/building Description

Operations Radiation 

type

Energy

Begin End (keV) Percentage

BGRR Reactor 1950 1969

Beta

Photon

>15

30-250

>250

100

25

75

HFBR/750 Reactor 1965 1999

BMRR/491 Reactor 1959 2000

MRC490/490A Radio-isotopes 1958 Present

RTF/490 Accelerators 1991 <2006 Electron

Photon

>15

>250

100

100

BLIP/931 Accelerator 1973 Present Electron

Photon

>15

30-250

>250

100

a

a

PET/490 Accelerator 1977 Present Electron

Photon

>15

>250

100

100

Cosmotron/same Cyclotron 1952 1966 Photon 30-250

>250

a

RHIC/same Accelerator 1999 Present a

AGS/same Synchrotron 1960 Present

Photon 30-250

> 250

a

a

AGS experimental area/912 Accelerator 1960 Present

AGS ring/913 1960 Present

AGS booster/942 1992 Present

200 MeV Linear Accelerator/ 

930

Accelerator 1972 Present

Tandem Van de Graaff/901 Accelerator 1970 Present

Tandem-to-Booster Beam 

Line/ Heavy Ion Beam 

tunnel

Accelerator 1991 Present

Accelerator Test Facility/ 

820

Accelerator 1989 Present Electron

Photon

>15

unknown

100

a

NSLS/725 Electron storage 1982 Present Electron

Photon

>15

30-250

100

100

Chemistry Dept./490A Accelerator Electron

Photon

>15

unknown

100

a

a. General guidance is to use 100% 30 – 250 keV for overestimates, 100% >250 keV for underestimates, 50% each for 

best estimate.

Table 6-3. Beta and photon energies and percentages (support facilities).

a

Process/building Description

Operations Radiation 

type

Energy

Begin End (keV) Percentage

NSRL NASA Space Lab 2003 Present Unknown Unknown Unknown

Tritium evaporator/802B Tritium processing 1995 Present Beta <15 avg. 100

WCF/811 Waste processing 1947 1987 Beta

Photon

>15

30-250

>250

100

25

75

TPL/801 Hot Laboratory 1951 Present

HWMF/444 Incinerator 1970s Early 

1990s

Chemistry Dept/555 & 490A Chemistry R&D 1966 Present

LEAF/555 Laser electron 

accelerator

1998 Present

Biology/463 Biological R&D
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Process/building Description

Operations Radiation 

type

Energy

Begin End (keV) Percentage

Oceanographic 

Sciences/318

Oceanographic 

R&D

Rad. Sciences/703, 703W HP & Env. Support

Nuclear Waste Mgt./830 Waste Handling

a. The isotopes involved ranged from H-3 to Au-198 resulting in beta, electron, and photon energies ranging from a few 

10s to several 1000s of keVs with wide ranging percentages.

6.7 NEUTRON RADIATIONS AND PERCENTAGES

Table 6-4 lists facilities with neutron radiations. They are reactors, accelerators, general laboratories, 

and medical facilities. The reactors are no longer in operation, the last being shut down in 2000.

Table 6-4. Facilities, neutron energies, percentages, and correction factors.

Facilities Source

Neutron

Energy

(MeV)

Dose

Fraction

ICRP 60a

Correction

Factor

BGRR, HFBR, BMRR Reactors 0.1–2.0 100% 1.91

Cosmotron, AGS, RHIC, NSLS Accelerators: Particle 

Interactionsb

0.1–2.0 75% 1.91

>2 25% 1.32

Medical, calibration, chemistry 

sources

252Cf 0.1–2 100% 1.91

AmBe, PuBe, 2–10 100% 1.32

Weighted Averagec 0.1–2 57% 1.66

>2 43%

a. ICRP (1991).

b. Select energy appropriate for facility and work activities.

c. Weighted average from alpha-neutron and spontaneous fission neutrons (ORAUT 2009).

Figures 6-1 and 6-2 show the neutron energy information upon which Table 6-4 is based. For Figure 

6-1, measurements in the thermal column were normalized such that the flux in the bare thermal 

column was 7 X 108 n/cm2

/sec. The thermal flux in the instrument tunnel was 2 X 109 n/cm2/sec. The 

converter plate was a U-235 plate used to convert the thermal beam into a fission spectrum.
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 84 of 115

Figure 6-1. Neutron Spectrum in the Thermal Column with the Converter Plate (left) and in the 

Instrument Tunnel (right).

Figure 6-2. Neutron Energy Spectrum at the 

Cosmotron (Sanna 1963).
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 85 of 115

All potential neutron exposures of >100 mrem annually used NTA film through 1995. From start up to 

1984, the NTA film was only processed if the density or darkening behind the cadmium (Cd) filter was 

greater than some predetermined value behind a photon equivalent, but neutron insensitive, filter 

(Faust 2010). This data could also be used to determine thermal neutron doses. Documentation of 

BNL’s method has not been found, but it is thought to be consistent with the practices of other sites 

(Caruthers 1964, Strom 1982). Caruthers (1964) indicates that a difference in exposure behind the 

Cd and neutron insensitive filters of 20 mR would be equivalent to about 10 mrem from thermal 

neutrons. BNL included the dose from thermal and slow neutrons in their fast neutron results. 

Starting in 1996, the Harshaw 8806 TLD along with CR-39 was used at the accelerators (Lane and 

Reciniello 2003; Sengupta 2000). Recorded neutron doses include both thermal and fast neutrons 

generated by the accelerators with energies >100 MeV (Sengupta 2000). The actual “dose of record” 

always included a QF of 10 starting from the first day of operation (Cowan ca. 1953). DOE is in the 

process of implementing ICRP Publication 60 (ICRP 1991) neutron weighting factors into the routine 

determination of the recorded neutron dose. For BNL, the date this is scheduled to begin is July 1, 

2010. Once this change is made, no adjustment in the recorded neutron dose (as described in the 

last column of Table 6-4) will be necessary, thereafter.

6.8 RECORDED DOSE PRACTICES

Recorded dose practices at BNL are given in Table 6-5 and include those provided by both the site 

and its vendor and any special dosimeter results (Sengupta 2000). Special dosimeters were worn if 

needed due to abnormal conditions, such as highly directional beams. If more than one dosimeter 

was required, the highest value was recorded. The regular dosimeter was not worn on these 

occasions (Sengupta 2000).

Table 6-5. Recorded dose practices.

1949–1984 Beta=Open window, mrema

Photon (P), mR

Fast Neutron (FN), mrem

Skin = OW+P

WB = P+ FN+ tritium 

1985–1995 Beta or nonpenetrating, mrem

Photon (P), mrem 

Fast Neutron (FN), mrem 

Skin = NP + P 

WB = NP + P + FN + tritium

1996–present Beta or nonpenetrating, mrem

Photon (P), mrem

Fast Neutron (FN), mrem 

Skin = NP + WB

WB = P + NP + FN + tritium

a. At startup, mrep was used interchangeably with mR, mrad, and later mrem.

b. Attached to either dosimeter when neutrons might be present.

6.9 INTERPRETATION OF REPORTED DOSES

Early personnel doses were reported in units of mrep for both penetrating (photon) and 

nonpenetrating (beta) if beta or low-energy electrons were present. High-energy beta or electrons 

were recorded as penetrating dose. The recorded total doses include the results from any special 

dosimeters worn for that exchange period. If it is necessary to obtain organ doses, dose 

reconstructors should use External Dose Reconstruction Implementation Guidelines (NIOSH 2007b). 

Table 6-6 lists the interpretation of the reported data by period. Reported doses have been corrected 

for background using site controls. The controls are dosimeters kept on site in locations used for the 

storage of personnel dosimeters. All issued dosimeters were stored on storage racks at positions 

throughout the site. Dosimeters were not to be taken home at the end of the shift, a practice 

beginning at the startup of the site and continuing at present (Sengupta 2000).
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Table 6-6. Interpretation of reported data.

Period

Reported 

quantity Description

Interpretation 

of zeros

Interpretation 

of blanks

Rollup of 

individual 

and annual 

data

Monitored/

unmonitored

Startup

–1985

Skin = mrad

WB =mR

Neutrons =

mrem

mrep (about 

startup)

mR, mrad, and 

mrem used 

interchangeably

MDL/2 times 

number of 

zeros

If no 

dosimeter for 

that period,

treat as 

unmonitored.

If special 

dosimeters 

were used, 

include 

results.

Only those

>10% of 

standards were 

monitored. 

Those entering 

controlled areas 

were issued 

visitors’ 

dosimeters.

1986–

present

Skin=mrem

WB= mrem

Neutrons=mrem

mrem used for 

all.

MDL/2 times 

number of 

zeros

If no 

dosimeter for 

that period, 

treat as 

unmonitored.

If special 

dosimeters 

were used, 

include 

results.

All personnel 

expected to be 

exposed to >100 

mrem in an 

exchange period 

were monitored.

6.10 ADJUSTMENTS TO RECORDED DOSE

Because most but not all penetrating photons are above 30 keV, it is suggested that an adjustment is 

necessary to account for the contribution to Hp(10) from the lower energy photons. It is estimated 

that a correction equal to 10% of the <250-keV values be made. Corrections are made for each 

energy group. Table 6-2 should be consulted for the applicable facility specific energy fractions. The 

NSLS photons are all considered to be in the lower energy group [measurements indicate a peak at 

75 keV (Preisig 1997)]. These adjustments also increase Hp(0.07) because that dose was always the 

sum of the open-window (OW) and WB doses. Table 6-7 lists the corrections. The fading correction 

for NTA film is about 9%/week (Meyer 1994), and the angular dependence correction factor is 1.3 

(Kathren 1965).

Table 6-7. Adjustments to recorded dose.

Period Facility Adjustment to reported dose

Photon energy Multiply by

Startup–1985 Reactors 30-250 keV 1.1

> 250 keV 1.0

1958–present MRC 490/490A 30-250 keV 1.1

> 250 keV 1.0

1982–present NSLS 30-250 keV 1.1

6.11 BIAS AND UNCERTAINTY

No site-specific data has been found about the bias and uncertainty for 1947 to 1984 when BNL used 

in-house film badges with NTA film. BNL did participate in some film badge performance tests in the 

mid-1960s, which can be used as an indication of the performance that was achieved.

In one test, 12 commercial film badges suppliers and two national laboratories were included in a 

blind intercomparison study of the accuracy and consistency of film badge readings that were 

exposed in March 1963. BNL was one of the national laboratories according to the 

acknowledgements. The national laboratories were not tested for consistency (precision), but their 

performance was consistent with the better-performing group of commercial processors. The study 

concluded that the accuracy that could be reasonably expected in practice, with a confidence limit of 

90%, was in the range of –50% to +200% for groups of badges exposed to X-rays, gamma rays, and 
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 87 of 115

mixtures thereof, over the exposure range of approximately 20 mR to 8,000 mR (Gorson, 

Suntharalingam, and Thomas 1965).

A 1967 study to determine film dosimeter performance criteria included participants from 15 AEC 

laboratories (Unruh et al. 1967). Although the results did not identify the laboratories, it is logical that 

BNL would have participated. Averaged over all radiation categories about 90% of the tests were 

passed by AEC contractors. The bias performance criterion was 90% to 110% (within 10%).

The uncertainty factors in Table 6-8 are proposed based on judgment of uncertainty of dosimeter 

performance in the site-specific fields for the period from start-up to 1984.

Table 6-8. Uncertainty factors.

Dosimeter Uncertainty Factora

Two-element film dosimeter (1947–1963) 1.4

Multi-element Film (1964–1984) 1.3

NTA Film (1954–1984) (2–14 MeV) 1.3

NTA Film (1954–1984) (0.1–2 MeV) 1.5

a Estimated variability of measured dose based on dosimeter 

technology as used in site radiation fields for long term workers.

In 1985, BNL contracted with R. S. Landauer Company for film badge services. The following table is 

a summary of the specifications from Landauer in 1988 (Schopfer 1988).

A few reports of quality control checks on Landauer’s performance were located. A report from 1986 

showed that Landauer was generally able to meet the “mean accuracy” specification (Schopfer 1986). 

In a 1988 report, the bias (average relative error) over the whole test range (up to 5000 mrem) for 

beta-gamma, CR-39, NTA and Lexan dosimeters was –0.09, –0.02, –0.17, and –0.03, respectively. It 

was noted, however, that the precision in this test failed to meet expectations (Schopfer 1988).

Table 6-9. Landauer gamma film, NTA film, and CR-39 

specifications.

Radiatio

n

Category

Source

Used

Range

(mrem)

Mean

Accuracy a

(mrem)

Precisio

n

(1 σ)

High 

Energy

Photon

Cs-137

50 20 20 mrem

500 50 10%

5000 500 10%

High 

Energy

X-ray

50 20 20 mrem

500 75 15%

5000 750 15%

Low 

Energy

X-ray

Am-241

15-20 keV

55-65 keV

50 20 20 mrem

500 75 25%

5000 750 25%

Fast

Neutron

NTA

AmBe

50 20 20 mrem

500 100 20%

5000 1000 15%

Fast

Neutron

CR-39

AmBe

50 20 20 mrem

500 100 20%

5000 1000 15%

a. Mean accuracy is the limit of the average difference of result of a 

group of dosimeters from the delivered exposure or dose.

In December 1995, BNL received DOELAP accreditation in external dosimetry (Loesch 1995). In 

1996, the Landauer film badge service was replaced by TLD system processing on site. The 
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 88 of 115

Landauer-supplied CR-39 and Lexan dosimeters were discontinued after June 1997. Bias and

uncertainty from 1996 on are expected to be consistent with DOELAP standards.
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 89 of 115

7.0

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

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

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

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

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

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

ATTRIBUTIONS AND ANNOTATIONS

[1] Potter, Eugene. M. H. Chew & Associates. Principal Health Physicist. November 2009.

Figure 2-2 was taken from a BNL document that has since been removed from the BNL web 

site. It was annotated by the section’s author with names and arrows that point out the major 

radioactive material and radiation-producing facilities. This figure is similar to figures that 

appear in a number of BNL site environmental reports, which may be consulted for 

comparison.

[2] Elyse Thomas. ORAUT Team. Principal Medical Dosimetrist. December 2009.

Review of claim file records from BNL. 

[3] Elyse Thomas. ORAUT Team. Principal Medical Dosimetrist. December 2009.

Review of claim file records from BNL. 

[4] Elyse Thomas. ORAUT Team. Principal Medical Dosimetrist. December 2009.

Review of claim file records from BNL. 

[5] Robert Morris. ORAUT Team. Principal Health Physicist. August 2006

Review of actual films at BNL.

[6] Robert Morris. ORAUT Team. Principal Health Physicist. August 2006

Review of actual films at BNL.

[7] Potter, Eugene. M. H. Chew & Associates. Principal Health Physicist. November 2009.

Conclusions were based on a review of the environmental reports listed in the References at 

the end of this revision.

[8] Potter, Eugene. M. H. Chew & Associates. Principal Health Physicist. November 2009.

The factor of 1.5 is suggested in ORAUT-PROC-0031 and is not accounted for separately in 

the tool used for dose reconstruction.

[9] Potter, Eugene. M. H. Chew & Associates. Principal Health Physicist. November 2009.

A reduction factor of 0.01 was selected as a conservative value after considering a number of 

different scenarios and typical values as input to atmospheric transport models.

[10] Potter, Eugene. M. H. Chew & Associates. Principal Health Physicist. November 2009.

This statement is based on professional judgment after reviewing the available data in a 

number of references.

[11] Potter, Eugene. M. H. Chew & Associates. Principal Health Physicist. November 2009.

The uncertainty range values quoted are based on professional judgment and experience.

[12] Potter, Eugene. M. H. Chew & Associates. Principal Health Physicist. November 2009.

These conclusions are based on the review of a number of whole body count results in 

individual dosimetry files.
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 90 of 115

[13] Potter, Eugene. M. H. Chew & Associates. Principal Health Physicist. November 2009.

These conclusions are based on the review of a number of whole body count results in 

individual dosimetry files, as well as other records and interviews.

[14] Potter, Eugene. M. H. Chew & Associates. Principal Health Physicist. November 2009.

These conclusions are based on the review of a number of bioassay data packages and whole 

body count records.

[15] Potter, Eugene. M. H. Chew & Associates. Principal Health Physicist. November 2009.

Based on a review of records and site interviews.

[16] Potter, Eugene. M. H. Chew & Associates. Principal Health Physicist. November 2009.

This statement is based on the historical monitoring thresholds.
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 91 of 115

REFERENCES

AEC (U.S. Atomic Energy Commission), 1949, Monthly Status and Progress Report for April 1949, 

New York Operations Office, New York, New York, May 10. [SRDB Ref ID: 6291, p. 93]

AEC (U.S. Atomic Energy Commission), 1949a, Monthly Status and Progress Report for January 

1949, New York Operations Office, New York, New York, February 7. [SRDB Ref ID: 6291, p. 

1]

AEC (U.S. Atomic Energy Commission), 1949b, Monthly Status and Progress Report for February 

1949, New York Operations Office, New York, New York, March 8. [SRDB Ref ID: 6291, p. 

25]

AEC (U.S. Atomic Energy Commission), 1949c, Monthly Status and Progress Report for March 1949, 

New York Operations Office, New York, New York, April 7. [SRDB Ref ID: 6291, p. 55]

AEC (U.S. Atomic Energy Commission), 1949e, Monthly Status and Progress Report for September 

1949, New York Operations Office, New York, New York, October 10. [SRDB Ref ID: 6291, 

p. 124]

AEC (U.S. Atomic Energy Commission), 1950, Monthly Status and Progress Report for December 

1949, New York Operations Office, New York, New York, January 5. [SRDB Ref ID: 6291, p. 

217]

AEC (U.S. Atomic Energy Commission), 1955, urinalysis report for Sample A8962, Form N.Y. 99, 

New York Operations Office, New York, New York, July 20. [SRDB Ref ID: 6274]

AEC (U.S. Atomic Energy Commission), 1957, urinalysis report for Sample B1300, Form N.Y. 99, 

New York Operations Office, New York, New York, December 16. [SRDB Ref ID: 6302, p. 5]

AEC (U.S. Atomic Energy Commission), 1958, urinalysis report for Sample A6352, Form N.Y. 99, 

New York Operations Office, New York, New York, October 14. [SRDB Ref ID: 6302, p. 8]

AEC (U.S. Atomic Energy Commission), 1961, urinalysis report for Sample D1780, Form N.Y. 99, 

New York Operations Office, New York, New York, August 28. [SRDB Ref ID: 6302, p. 28]

AEC (U.S. Atomic Energy Commission), 1962, urinalysis report for Sample D607, Form N.Y. 99, New 

York Operations Office, New York, New York, August 14. [SRDB Ref ID: 6302, p. 32]

AEC (U.S. Atomic Energy Commission), 1963, urinalysis report for Sample D756, Form N.Y. 99, New 

York Operations Office, New York, New York, April 16. [SRDB Ref ID: 6302, p. 35]

Astarita Associates, 1998–2004, collection of reports on Patient Radiographic Skin Doses, Smithtown, 

New York. [SRDB Ref ID: 22528]

Bishop, W. H. Jr., 1958, “Progress Report,” memorandum to L. Gemmell, Associated Universities, 

Brookhaven National Laboratory, Upton, New York, April 8. [SRDB Ref ID: 22919, p. 7]

BNL (Brookhaven National Laboratory), undated, excerpt from HFBR Safety Analysis Report: 

Radioactive Wastes and Effluents, Upton, New York. [SRDB Ref ID: 22464]

BNL (Brookhaven National Laboratory), 1951, Health Physics Summary for August, 1951, Associated 

Universities, Upton, New York, August. [SRDB Ref ID: 22848, p. 8]
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 92 of 115

BNL (Brookhaven National Laboratory), 1952, Health Physics Summary for May 1952, Associated 

Universities, Upton, New York, May. [SRDB Ref ID: 22915, p. 6]

BNL (Brookhaven National Laboratory), 1963, Heath Physics and Safety Summary for May 1963, 

Associated Universities, Upton, New York, Undated. [SRDB Ref ID: 6517, p. 96]

BNL (Brookhaven National Laboratory), 1972, 1971 Environmental Monitoring Report, Associated 

Universities, Upton, New York, July. [SRDB Ref ID: 6356]

BNL (Brookhaven National Laboratory), 1973, Safety Policy Manual, Brookhaven National Laboratory, 

Upton, New York, April. [SRDB Ref ID: 22479]

BNL (Brookhaven National Laboratory), 1993, “8.7 Internal Monitoring,” RD APM 8.7, Rev. 0, 

Associated Universities, Upton, New York, August 13. [SRDB Ref ID: 55393]

BNL (Brookhaven National Laboratory), 1995, “Radiological Health Support Operations,” Chapter 5, 

Radiological Control Manual, Associated Universities, Brookhaven National Laboratory, Upton, 

New York, November 20. [SRDB Ref ID: 22455]

BNL (Brookhaven National Laboratory), 1997a, Review of Potential Environmental Release Points, 

Draft, Directors Office Building 701/BGRR, Associated Universities, Upton, New York. [SRDB 

Ref ID: 77713]

BNL (Brookhaven National Laboratory), 1997b, Review of Potential Environmental Release Points, 

Directors Office Buildings 701, 704, 708, 709, Rev. 1, Associated Universities, Upton, New 

York. [SRDB Ref ID: 77712]

BNL (Brookhaven National Laboratory), 1997c, Review of Potential Environmental Release Points, 

Department/Division -Reactor Building/Facility -750/HFBR, 707A/Pumphouse, 707B/Water 

Treatment House & Cooling Towers, Associated Universities, Upton, New York June 2. 

[SRDB Ref ID: 77716]

BNL (Brookhaven National Laboratory), 1997d, Review of Potential Environmental Release Points, 

Medical Department/490, Rev. 1, Associated Universities, Upton, New York. [SRDB Ref ID: 

77702]

BNL (Brookhaven National Laboratory), 1997e, Review of Potential Environmental Release Points, 

Reactor Division, Building 491/Brookhaven Medical Research Reactor, Associated 

Universities, Upton, New York, July. [SRDB Ref ID: 77703]

BNL (Brookhaven National Laboratory), 1997f, Review of Potential Environmental Release Points, 

Chemistry Department, Building 906, Associated Universities, Upton, New York. [SRDB Ref 

ID: 77720]

BNL (Brookhaven National Laboratory), 1997g, Review of Potential Environmental Release Points, 

RHIC Project, Building 902 Complex, Associated Universities, Brookhaven National 

Laboratory, Upton, New York, June 18. [SRDB Ref ID: 77719]

BNL (Brookhaven National Laboratory), 1997h, Review of Potential Environmental Release Points,

AGS Department, Building Forms: Level I, II & III, Associated Universities, Upton, New York, 

June. [SRDB Ref ID: 77723]
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 93 of 115

BNL (Brookhaven National Laboratory), 1997i, Review of Potential Environmental Release Points, 

RHIC Project, Building 960 Complex, Associated Universities, Brookhaven National 

Laboratory, Upton, New York, June 18. [SRDB Ref ID: 77722]

BNL (Brookhaven National Laboratory), 1997j, Review of Potential Environmental Release Points, 

Medical Department, Building 931, Associated Universities, Upton, New York. [SRDB Ref ID: 

77721]

BNL (Brookhaven National Laboratory), 1997k, Review of Potential Environmental Release Points, 

Chemistry Department, Building 901, Associated Universities, Brookhaven National 

Laboratory, Upton, New York, May. [SRDB Ref ID: 77718]

BNL (Brookhaven National Laboratory), 1997l, NSLS Building 725 Final Facility Review Report, 

NSLS, Associated Universities, Upton, New York. [SRDB Ref ID: 77714]

BNL (Brookhaven National Laboratory), 1997m, NSLS Building 729 Final Facility Review Report, 

NSLS, Associated Universities, Upton, New York, June 17. [SRDB Ref ID: 77715]

BNL (Brookhaven National Laboratory), 1997n, Review of Potential Environmental Release Points, 

RHIC Project, RHIC Complex, Associated Universities, Brookhaven National Laboratory, 

Upton, New York, June. [SRDB Ref ID: 77724]

BNL (Brookhaven National Laboratory), 1997o, Review of Potential Environmental Release Points, 

Medical Department/Division, Building/Facility: Building 801, Rev. 2, Associated Universities, 

Upton, New York, July 25. [SRDB Ref ID: 71622]

BNL (Brookhaven National Laboratory), 1997p, Review of Potential Environmental Release Points, 

Department of Advanced Technology, Building 830, Associated Universities, Upton, New York, 

June. [SRDB Ref ID: 77717]

BNL (Brookhaven National Laboratory), 1997q, Review of Potential Environmental Release Points, 

Chemistry Department, Building 555, Associated Universities, Upton, New York, June. [SRDB 

Ref ID: 77707]

BNL (Brookhaven National Laboratory), 1997r, Review of Potential Environmental Release Points, 

Physics Department, Building 510 Including Trailers, Associated Universities, Upton, New 

York, September. [SRDB Ref ID: 77704]

BNL (Brookhaven National Laboratory), 1997s, Review of Potential Environmental Release Points, 

RHIC Project Buildings 620, 820, 835, 836, Associated Universities, Upton, New York, June. 

[SRDB Ref ID: 77708]

BNL (Brookhaven National Laboratory), 1997t, Review of Potential Environmental Release Points, 

Physics Department/Division Central Shops, Division Building/Facility 530, Associated 

Universities, Upton, New York, June. [SRDB Ref ID: 77705]

BNL (Brookhaven National Laboratory), 1997u, Review of Significant Environmental Release Issues, 

Instrument Division, Building 535, Associated Universities, Upton, New York, June. [SRDB 

Ref ID: 77706]

BNL (Brookhaven National Laboratory), 1997v, Review of Potential Environmental Release Points, 

Safety & Environmental Protection Division Bldg. 855, 860, 865 and 870, Associated 

Universities, Upton, New York, June. [SRDB Ref ID: 77725]
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 94 of 115

BNL (Brookhaven National Laboratory), 1997w, Review of Potential Environmental Release Points, 

Safety & Environmental Protection/Plant Engineering, Building 650-650T/Radiological Waste 

Decontamination & Hot Laundry Facility, Associated Universities, Upton, New York, August. 

[SRDB Ref ID: 77709]

BNL (Brookhaven National Laboratory), 1999, 1998 Site Environmental Report, BNL-52572, 

Brookhaven Science Associates, Upton, New York, September. [SRDB Ref ID: 6478]

BNL (Brookhaven National Laboratory), 1999, Tritium Dose to HFBR Personnel, July-December 

1999, Brookhaven Science Associates, Upton, New York. [SRDB Ref ID: 22425]

BNL (Brookhaven National Laboratory), 2000, 1999 Site Environmental Report, BNL-52585, 

Brookhaven Science Associates, Upton, New York, August. [SRDB Ref ID: 6480] 

BNL (Brookhaven National Laboratory), 2001a, 2000 Site Environmental Report, BNL-52626, 

Brookhaven Science Associates, Upton, New York, September. [SRDB Ref ID: 6484] 

BNL (Brookhaven National Laboratory), 2001b, Technical Basis for the BNL Internal Dosimetry 

Program, Rev. 3, Brookhaven Science Associates, Upton, New York, March 29. [SRDB Ref 

ID: 22519]

BNL (Brookhaven National Laboratory), 2002, 2001 Site Environmental Report, BNL-52671, 

Brookhaven Science Associates, Upton, New York, September. [SRDB Ref ID: 33177]

BNL (Brookhaven National Laboratory), 2003, 2002 Site Environmental Report, BNL-71521-2003, 

Brookhaven Science Associates, Upton, New York, October. [SRDB Ref ID: 33167]

BNL (Brookhaven National Laboratory), 2004, 2003 Site Environmental Report, BNL-73160-2004, 

Brookhaven Science Associates, Upton, New York. [SRDB Ref ID: 33169]

BNL (Brookhaven National Laboratory), 2005, 2004 Site Environmental Report, BNL-74909-2005, 

Brookhaven Science Associates, Upton, New York, October. [SRDB Ref ID: 33171]

BNL (Brookhaven National Laboratory), 2009a, “Image Library,” Brookhaven Science Associates, 

Upton, New York, March 9. [SRDB Ref ID: 79734]

BNL (Brookhaven National Laboratory), 2009b, “The Physics of RHIC,” Brookhaven Science 

Associates, Upton, New York, March 9. [SRDB Ref ID: 79735]

Bond, V. P., 1963, “Medical Examinations of Reactor Operators,” memorandum to C. Williams, 

Associated Universities, Brookhaven National Laboratory, Upton, New York, January 29. 

[SRDB Ref ID: 53887]

Brodksy, A., 1964, minutes of meetings. [SRDB Ref ID: 17344]

Carter, E. N., 1975a, “Progress Report May/June, 1975,” memorandum to C. W. Flood, Associated 

Universities, Brookhaven National Laboratory, Upton, New York, September 24. [SRDB Ref 

ID: 22832, p. 8]

Carter, E. N., 1975b, “Progress Report January/February, 1975,” memorandum to C. W. Flood, 

Associated Universities, Brookhaven National Laboratory, Upton, New York, April 3. [SRDB 

Ref ID: 22832, p. 16]
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 95 of 115

Carter, J., 1978, “Radiation Protection Guidance to Federal Agencies for Diagnostic X Rays,” Federal 

Register, volume 43, number 22, February 1. [SRDB Ref ID: 8141]

Carter, N., 1978, “X-Ray Exposures,” memorandum to R. Fairchild, Associated Universities, 

Brookhaven National Laboratory, Upton, New York, October 11. [SRDB Ref ID: 53909]

Caruthers, L. T. and Story, E. J., 1964, “Capture Gammas for Neutron Dosimetry with a Film Badge,” 

Health Physics, volume 10, September, pp. 667-672

Conrad, R. D., 1948, “Medical Department X-Ray Exposure Records.” [SRDB Ref ID 55355]

Cooley, R. C, 1967, “Personnel Radiation Exposures from Medical X-rays – Savannah River Plant –

Part II,” memorandum to E. C. Morris, July 11. [SRDB Ref ID: 68124]

Cowan, F. P., 1948, “X-Ray Exposure Records,” memorandum to R. D. Conrad, Associated 

Universities, Brookhaven National Laboratory, Upton, New York, April 28. [SRDB Ref ID: 

55356].

Cowan, F. P., 1948, “X-Ray Exposure Records,” memorandum to R. D. Conrad, Associated 

Universities, Brookhaven National Laboratory, Upton, New York, April 26. [SRDB Ref ID: 

55356].

Cowan, F. P., 1957a, “Radiation Safety in a Research Laboratory,” Associated Universities, 

Brookhaven National Laboratory, Upton, New York, May. [SRDB Ref ID: 22501]

Cowan, F. P., 1957b, “P32 Spill of April 23, 1957,” memorandum to J. B. H. Kuper, Associated 

Universities, Brookhaven National Laboratory, Upton, New York, June 7. [SRDB Ref ID: 

6319, p. 85]

Cowan, F. P., 1963, “Personnel Dosimetry of Very-High Energy Radiations,” Symposium On 

Personnel Dosimetry Techniques For External Radiation, Madrid, 1st - 5th April 1963, 

European Nuclear Energy Agency and Junta de Energia Nuclear, p. 273. [SRDB Ref ID: 

22498]

Cowan, F. P., ca. 1953, “A Preliminary Report on Health Physics Problems at the Brookhaven 

Alternating Gradient Synchrotron,” Colloque sur la Protection Aupres des Grands 

Accelerateurs, p. 143. [SRDB Ref ID: 22497]

Cowan, F. P., undated, Health Physics Program for the Brookhaven National Laboratory 

Synchrotrons, Associated Universities, Brookhaven National Laboratory, Upton, New York. 

[SRDB Ref ID: 22490]

Cowan, F. P., L. B. Farabee, and R. A. Love, 1952, “Health Physics and Medical Aspects of a 

Strontium 90 Inhalation Incident,” The American Journal of Roentgenology and Radium 

Therapy, volume 67, number 5. pp. 805–809. [SRDB Ref ID: 22543]

Cowan, F. P., and J. S. Handloser, 1953, Health Physics Program for the Brookhaven Cosmotron, 

BNL-264, Associated Universities, Brookhaven National Laboratory, Upton, New York, 

November. [SRDB Ref ID: 22516]

Cowan, F. P., L. F. Phillips, and R. J. King, 1963, Some Dosimetry Problems of the Alternating 

Gradient Synchrotron (AGS), BNL 7151, Associated Universities, Brookhaven National 

Laboratory, Upton, New York, July 1. [SRDB Ref ID: 22492]
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 96 of 115

Cowan, F. P., and J. Weiss, 1957, “Bioassay Data and Analysis Relating to the P32 Spill of April 23, 

1957,” memorandum to R. A. Love, Associated Universities, Brookhaven National Laboratory, 

Upton, New York, June 7. [SRDB Ref ID: 6319, p. 80]

Crease, R. P., 1999, Making Physics, A Biography of Brookhaven National Laboratory, 1946–1972, 

University of Chicago Press, Chicago, Illinois. [SRDB Ref ID: 23541]

Day, L. E., and J. R. Naidu, editors, 1983, 1982 Environmental Monitoring Report, BNL 51697, 

Associated Universities, Brookhaven National Laboratory, April. [SRDB Ref ID: 6405]

Day, L. E., and J. R. Naidu, editors, 1984, 1983 Environmental Monitoring Report, BNL 51827, 

Associated Universities, Brookhaven National Laboratory, Upton, New York, June. [SRDB Ref 

ID: 6409]

Day, L. E., R. P. Miltenberger, and J. R. Naidu, editors, 1985, 1984 Environmental Monitoring Report, 

BNL 51884, Associated Universities, Brookhaven National Laboratory, Upton, New York, April. 

[SRDB Ref ID: 6414]

Day, L. E., R. P. Miltenberger, and J. R. Naidu, editors, 1986, 1985 Environmental Monitoring Report, 

BNL 51993, Associated Universities, Brookhaven National Laboratory, Upton, New York, April. 

[SRDB Ref ID: 6426]

DOE (U.S. Department of Energy), 1999, Draft Environmental Impact Statement for the High Flux 

Beam Reactor Transition Project at the Brookhaven National Laboratory, Brookhaven National 

Laboratory, Upton, New York, Office of Basic Energy Sciences, Germantown, Maryland, April. 

[SRDB Ref ID: 22517]

Eisenbud, M., 1957, “Explosion at Brookhaven National Laboratory,“ memorandum to K. E. Fields, 

U.S. Atomic Energy Commission, New York Operations Office, New York, New York, June 7. 

[SRDB Ref ID: 6319, p. 31]

ERDA (Energy Research and Development Administration), 1977, Final Environmental Impact 

Statement, Brookhaven National Laboratory, Upton, New York, ERDA-1540, Washington D.C., 

July. [SRDB Ref ID: 6384]

Faust, L. G., 2010, “NTA Film Usage,” documented communication with H. Kahnhauser., Oak Ridge 

Associated Universities Team, Oak Ridge, Tennessee. March 26. [SRDB Ref ID: 80141]

Flood, C., 1976, “Health Protection and Health Surveillance,” memorandum to V. Pai, Associated 

Universities, Brookhaven National Laboratory, Upton, New York, May 17. [SRDB Ref ID: 

55020].

Gorson, R. O., N. Suntharalingam, and J. W. Thomas, 1965, “Results of a Film-Badge Reliability 

Study,” unknown journal, volume 84, February, pp. 333–346. [SRDB Ref ID: 15022]

Handloser, J. S., ca. 1959, Stray Radiation Dosimetry Around Pulsed Accelerators, BNL 3519, 

Associated Universities, Brookhaven National Laboratory, Upton, New York. [SRDB Ref ID: 

22500]

Handloser, J. S., and R. A. Love, 1951, “Radiation Doses from Diagnostic X-Ray Studies,” Radiology, 

volume 57, pp. 252–254. [SRDB Ref ID: 13354]
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 97 of 115

Hardy, H.; Hodges, H.; and Wolf, B. S., 1948, “Tentative Program for Control of Beryllium Hazard in 

AEC Contracts,” memorandum with attachment to S. Warren (Washington, D.C.), U.S. Atomic 

Energy Commission, New York Operations Office, August 4. [SRDB Ref ID: 11318, p. 74]

Haworth, L. J., 1960, untitled letter to E. L. Van Horn (U.S. Atomic Energy Commission, Brookhaven 

Area Office), Associated Universities, Brookhaven National Laboratory, Upton, New York, 

September 29. [SRDB Ref ID: 6271]

Holeman, G. R., 1994, Personnel Monitoring External Dosimetry Quality Assurance Plan, PM-QAP001, Rev. 0, Associated Universities, Brookhaven National Laboratory, Upton, New York, 

December 27. [SRDB Ref ID: 22787]

Holeman, G. R., 1999, “A Document Describing the External and Internal Dosimetry Programs,” email with attachment to H. F. Kahnhauser, Brookhaven Science Associates, Brookhaven 

National Laboratory, Upton, New York, March 24. [SRDB Ref ID: 22477]

Hull, A. P., 1963, 1962 Environmental Radiation Levels at Brookhaven National Laboratory, BNL 807, 

Associated Universities, Brookhaven National Laboratory, Upton, New York, May. [SRDB Ref 

ID: 6330]

Hull, A. P., 1964, 1963 Environmental Radiation Levels at Brookhaven National Laboratory, BNL 915, 

Associated Universities, Brookhaven National Laboratory, Upton, New York, November. 

[SRDB Ref ID: 6333]

Hull, A. P., 1966, 1964 Environmental Radiation Levels at Brookhaven National Laboratory, BNL

50001, Associated Universities, Brookhaven National Laboratory, Upton, New York, June. 

[SRDB Ref ID: 6334]

Hull, A. P., editor, 1973, 1972 Environmental Monitoring Report, BNL 17874, Associated Universities, 

Brookhaven National Laboratory, Upton, New York, May. [SRDB Ref ID: 6359]

Hull, A. P., and J. A. Ash, editors, 1974, 1973 Environmental Monitoring Report, BNL 18625, 

Associated Universities, Brookhaven National Laboratory, Upton, New York, March. [SRDB 

Ref ID: 6360]

Hull, A. P., and J. A. Ash, editors, 1975, 1974 Environmental Monitoring Report, BNL-19977, 

Associated Universities, Brookhaven National Laboratory, Upton, New York, April. [SRDB Ref 

ID: 6362]

Hull, A. P., and J. A. Ash, editors, 1976, 1975 Environmental Monitoring Report, BNL-21320,

Associated Universities, Brookhaven National Laboratory, Upton, New York, March. [SRDB 

Ref ID: 6369]

Hull, A. P., and J. T. Gilmartin, 1967, 1965 Environmental Radiation Levels at Brookhaven National 

Laboratory, BNL 50093, Associated Universities, Brookhaven National Laboratory, Upton, 

New York, September. [SRDB Ref ID: 6340]

Hull, A. P., and J. T. Gilmartin, 1969, 1966 Environmental Radiation Levels at Brookhaven National 

Laboratory, BNL 50196, Associated Universities, Brookhaven National Laboratory, Upton, 

New York, September. [SRDB Ref ID: 6353]
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 98 of 115

ICRP (International Commission on Radiological Protection), 1975, Reference Man: Anatomical, 

Physiological and Metabolic Characteristics, Publication 23, Pergamon Press, Oxford, 

England.

ICRP (International Commission on Radiological Protection), 1983, Protection of the Patient in 

Diagnostic Radiology, Publication 34, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1991, 1990 Recommendation of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England.

Kathern R. L., Prevo, C. T., and Block, S., 1965, “Angular Dependence Of Eastman Type A (NTA) 

Personnel Monitoring Film,” Health Physics, volume 11, October, pp. 1067–1069.

Lane, G., and R. Reciniello, 2004, “Brookhaven National Laboratory, e-mail - Georgette Lane and 

Richard Reciniello 12/22/2003 and 3/31/2004.” [SRDB Ref ID: 18040]

Lee, R. J., J. R. Naidu, G. L. Schroeder, D. E. Paquette, S. Scarpitta, R. Howe, W. Dorsch, A. Bou, 

and W. Gunther, 1999, Brookhaven National Laboratory Site Report for Calendar Year 1997, 

Brookhaven Science Associates, Brookhaven National Laboratory, Upton, New York, 

February. [SRDB Ref ID: 6471]

Lessard, E. T., C. Schaefer, and R. Karol, 2001, Technical Basis for Bioassay Requirements, ColliderAccelerator Department, Brookhaven Science Associates, Brookhaven National Laboratory, 

Upton, New York, January 22. [SRDB Ref ID: 22428, p. 10]

Loesch, R. M., 1995, “Laboratory Accreditation of Brookhaven National Laboratory (BNL) Personnel 

Dosimetry,”, memorandum to C. L. Nealy (Brookhaven Area Office), U.S. Department of 

Energy, Office of Worker Protection Programs and Hazards Management, Washington, D.C., 

December 12. [SRDB Ref ID: 22457]

Love, R. A., 1957, untitled letter to B. A. Brundage (Atomics International), Associated Universities, 

Brookhaven National Laboratory, Upton, New York, November 13. [SRDB Ref ID: 53880]

Love, R. A., 1959, untitled letter to T. M. Fleidner (Universitat Heidelberg), Associated Universities, 

Brookhaven National Laboratory, Upton, New York, September 25. [SRDB Ref ID: 53618]

Love, R. A., 1963, untitled letter to C. L. Gilbert (Columbia University), Associated Universities, 

Brookhaven National Laboratory, Upton, New York, November 19. [SRDB Ref ID: 53891]

Love, R. A., 1964, untitled letter to H. J. Thompson (University of Minnesota), Associated Universities, 

Brookhaven National Laboratory, Upton, New York, June 8. [SRDB Ref ID: 53893]

Meinhold, C. B., and A. P. Hull, 1998, Radiological Environmental Monitoring Report for Brookhaven 

National Laboratory, 1967-1970, BNL 52551, Brookhaven Science Associates, Brookhaven 

National Laboratory, Upton, New York, October. [SRDB Ref ID: 6497]

Meinhold, C. B., and A. F. Meinhold, 2001, Radiological Emissions and Environmental Monitoring for 

Brookhaven National Laboratory, 1947 - 1961, P. D. Bond, editor, Brookhaven Science 

Associates, Brookhaven National Laboratory, Upton, New York, May. [SRDB Ref ID: 6495]
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 99 of 115

Meyer, H. E. (compiler), 1994, History of Personnel External Dosimetry Program at the Dayton Project 

and Mound Laboratory 1946-1993, Mound Laboratory, Miamisburg, Ohio. [SRDB Ref ID: 

3276]

Michel, L., 2003, “MDA Determination for the WBC,” memorandum to R. Reciniello, Brookhaven 

Science Associates, Brookhaven National Laboratory, Upton, New York, November 26. 

[SRDB Ref ID: 17882]

Michel, L., 2004, “MDA Determination for the WBC,” memorandum to R. Reciniello, Brookhaven 

Science Associates, Brookhaven National Laboratory, Upton, New York, July 15. [SRDB Ref 

ID: 22481, p. 4]

Michel, L., 2005, “MDA Determination for the WBC,” memorandum to R. Reciniello, Brookhaven 

Science Associates, Brookhaven National Laboratory, Upton, New York, May 17. [SRDB Ref 

ID: 22481, p. 6]

Michel, L., and C. Sun, 2003, “MDA Determination for the WBC,” memorandum to R. Reciniello, 

Brookhaven Science Associates, Brookhaven National Laboratory, Upton, New York, March 

28. [SRDB Ref ID: 22481, p. 2]

Miltenberger, R. P., 1978, “I-126 Release of May 16, 1978 from BLIF Facility,” memorandum to file, 

Associated Universities, Brookhaven National Laboratory, Upton, New York, August 18. 

[SRDB Ref ID: 58853]

Miltenberger, R. P., 1981a, “Whole Body Counting Results for Building 750,” memorandum to A. 

Humm, Associated Universities, Brookhaven National Laboratory, Upton, New York, January 

8. [SRDB Ref ID: 50910]

Miltenberger, R. P., 1981b, “Whole-Body Counting Results,” memorandum to A. Lukas, Associated 

Universities, Brookhaven National Laboratory, Upton, New York, December 21. [SRDB Ref 

ID: 58848]

Miltenberger, R. P., 1989, “Radiological Results of Thyroid Counting Performed for Medical 

Department Personnel on June 1, 1989,” memorandum to N. Carter, Associated Universities, 

Brookhaven National Laboratory, Upton, New York, June 7. [SRDB Ref ID: 22420, p. 9]

Miltenberger, R. P., N. A. Greenhouse, and E. T. Lessard, 1980, “Whole Body Counting Results from 

1974 to 1979 for Bikini Island Residents,” Health Physics, volume 39, September, pp. 395–

407.

Miltenberger, R. P., and E. T. Lessard, 1987, “Suspected Uptake of Xenon and Iodine by [Privacy Act 

Information],” memorandum to file, Associated Universities, Brookhaven National Laboratory, 

Upton, New York, January 14. [SRDB Ref ID: 58996]

Miltenberger, R. P., B. A. Royce, and J. R. Naidu, 1992, Brookhaven National Laboratory Site Report 

for Calendar Year 1990, BNL 52322, Associated Universities, Brookhaven National 

Laboratory, Upton, New York, January. [SRDB Ref ID: 6445]

Miltenberger, R. P., B. A. Royce, and J. R. Naidu, editors, 1987, 1986 Environmental Monitoring 

Report, BNL 52088, Associated Universities, Brookhaven National Laboratory, Upton, New 

York, June. [SRDB Ref ID: 6429]
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 100 of 115

Miltenberger, R. P., B. A. Royce, and J. R. Naidu, editors, 1988, 1987 Environmental Monitoring 

Report, BNL 52152, Associated Universities, Brookhaven National Laboratory, Upton, New 

York, April. [SRDB Ref ID: 6433]

Miltenberger, R. P., B. A. Royce, and J. R. Naidu, editors, 1989, Brookhaven National Laboratory Site 

Report for Calendar Year 1988, BNL-52207, Associated Universities, Brookhaven National 

Laboratory, Upton, New York, April. [SRDB Ref ID: 6437]

Miltenberger, R. P., B. A. Royce, S. S. Chalasani, D. Morganelli, and J. R. Naidu, 1990, Brookhaven 

National Laboratory Site Report for Calendar Year 1989, BNL 52264, Associated Universities, 

Brookhaven National Laboratory, Upton, New York, December. [SRDB Ref ID: 6441]

Miltenberger, R. P., and J. Steimers, 1984, “Bioassay Results,” memorandum to A. Lukas, Associated 

Universities, Brookhaven National Laboratory, Upton, New York, June 11. [SRDB Ref ID: 

22817]

Morris, R. L., 2006a, “BNL Occupational X-Ray: Interview with [Name Redacted], RT, X-Ray 

Technician (retired), Brookhaven National Laboratory,” documented communication, Oak 

Ridge Associated Universities Team, Oak Ridge, Tennessee, March 3. [SRDB Ref ID: 

22720]

Morris, R. L., 2006b, “BNL Occupational X-Ray: Interview with [Name Redacted], RT, X-Ray 

Technician, Brookhaven National Laboratory,” documented communication, Oak Ridge 

Associated Universities Team, Oak Ridge, Tennessee, March 2. [SRDB Ref ID: 22453]

Morris, R. L., 2006c, “BNL Occupational X-Ray: Interview with [Name Redacted], RT, X-Ray 

Technician (retired), Brookhaven National Laboratory,” documented communication, Oak 

Ridge Associated Universities Team, Oak Ridge, Tennessee, March 28. [SRDB Ref ID: 

22722]

Morris, R. L., 2006d, “Historical Info on OMC X-Ray Policies,” documented communication, Oak 

Ridge Associated Universities Team, Oak Ridge, Tennessee, March 28. [SRDB Ref ID: 

22721]

Murray, B., 1993, “Results of Annual Routine Whole Body Counts Performed of HFBR Personnel from 

7/16 through 9/16/1993,” memorandum to R. Reciniello, Associated Universities, Brookhaven 

National Laboratory, Upton, New York, October 19. [SRDB Ref ID: 22792, p. 21]

Murray, B., 1999, “A priori Minimum Detectable Activities of Representative Nuclides for the Accuscan 

II In Vivo Counters Using the Total Body Geometry: (MDA Workaround),” memorandum to G. 

Holeman, Brookhaven Science Associates, Brookhaven National Laboratory, Upton, New 

York, August 25. [SRDB Ref ID: 22423, p. 4]

Musolino, S. V., 2000, Technical Basis for Bioassay Requirements, BGRR Decommissioning Project, 

Brookhaven Science Associates, Brookhaven National Laboratory, Upton, New York, 

September 27. [SRDB Ref ID: 22428, p. 2]

Naidu, J. R., editor, 1977, 1976 Environmental Monitoring Report, BNL-22627, Associated 

Universities, Brookhaven National Laboratory, Upton, New York, April. [SRDB Ref ID: 6379]

Naidu, J. R., editor, 1979, 1978 Environmental Monitoring Report, BNL 51031, Associated 

Universities, Brookhaven National Laboratory, Upton, New York, April. [SRDB Ref ID: 6387]
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 101 of 115

Naidu, J. R., editor, 1980, 1979 Environmental Monitoring Report, BNL 51252, Associated 

Universities, Brookhaven National Laboratory, Upton, New York, April. [SRDB Ref ID: 6390]

Naidu, J. R., and L. L. Olmer, editors, 1981, 1980 Environmental Monitoring Report, BNL 51417, 

Associated Universities, Brookhaven National Laboratory, Upton, New York, April. [SRDB Ref 

ID: 6396]

Naidu, J. R., and L. L. Olmer, editors, 1982, 1981 Environmental Monitoring Report, BNL 51724, 

Associated Universities, Brookhaven National Laboratory, Upton, New York, April. [SRDB Ref 

ID: 6399]

Naidu, J. R., D. E. Paquette, G. L. Schroeder, and R. J. Lee, editors, 1996, Brookhaven National 

Laboratory Site Report for Calendar Year 1995, BNL-52522, Associated Universities, 

Brookhaven National Laboratory, Upton, New York, December. [SRDB Ref ID: 6466]

Naidu, J. R., and B. A. Royce, editors, 1994, Brookhaven National Laboratory Site 1976 

Environmental Monitoring Report for Calendar Year 1993, BNL-52432, Associated 

Universities, Brookhaven National Laboratory, Upton, New York, May. [SRDB Ref ID: 6459]

Naidu, J. R., and B. A. Royce, editors, 1995, Brookhaven National Laboratory Site Report for 

Calendar Year 1994, BNL-52490, Associated Universities, Brookhaven National Laboratory, 

Upton, New York, May. [SRDB Ref ID: 6461]

Naidu, J. R., B. A. Royce, and R. P. Miltenberger, 1992, Brookhaven National Laboratory Site 

Environmental Report for Calendar Year 1991, BNL 52347, Associated Universities, 

Brookhaven National Laboratory, Upton, New York, September. [SRDB Ref ID: 6451]

Naidu, J. R., B. A. Royce, and R. P. Miltenberger, 1993, Site Environmental Report for Calendar Year 

1992, BNL-52411, Associated Universities, Brookhaven National Laboratory, Upton, New 

York, May. [SRDB Ref ID: 6455]

NCRP (National Council on Radiation Protection and Measurements), 1976, Structural Shielding 

Design and Evaluation for Medical Use of X Rays and Gamma Rays of Energies up to 10 

MeV, Report 49, Bethesda, Maryland.

NCRP (National Council on Radiation Protection and Measurements), 1989, Medical X-Ray, Electron 

Beam and Gamma-Ray Protection for Energies up to 50 MeV, 102, Bethesda, Maryland.

Nelson, R., 1973, “Calibration of Doses from the Diagnostic X-Ray Machines,” memorandum to C. W. 

Flood, Associated Universities, Brookhaven National Laboratory, Upton, New York, August 13. 

[SRDB Ref ID: 55059]

NIOSH (National Institute for Occupational Safety and Health), 2007a, Radiation Exposures Covered 

for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, OCAS-IG-003, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, November 5.

NIOSH (National Institute for Occupational Safety and Health), 2007b, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 3, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, November 21.
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 102 of 115

NIOSH (National Institute for Occupational Safety and Health), 2002, Internal Dose Reconstruction 

Implementation Guideline, OCAS-IG-002, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, August 1.

O’Connell, E. J. III, 1975, “Progress Report - January, 1975,” memorandum to C. W. Flood, 

Associated Universities, Brookhaven National Laboratory, Upton, New York, February 20. 

[SRDB Ref ID: 22831, p. 31]

ORAUT (Oak Ridge Associated Universities Team), 2004, Technical Basis Document for the Oak 

Ridge National Laboratory -- Occupational Environmental Dose, ORAUT-TKBS-0012-4, Rev. 

00, Oak Ridge, Tennessee, May 7.

ORAUT (Oak Ridge Associated Universities Team), 2005, Dose Reconstruction from Occupationally 

Related Diagnostic X-Ray Procedures, ORAUT-OTIB-0006, Rev. 03 PC-1, Oak Ridge 

Tennessee, December 21. 

ORAUT (Oak Ridge Associated Universities Team), 2006, Atomics International – Occupational 

External Dose, ORAUT-TKBS-0038-6, Rev. 01, Oak Ridge Tennessee, November 16.

ORAUT (Oak Ridge Associated Universities Team), 2006, Estimating the Maximum Plausible Dose to 

Workers at Atomic Weapons Employer Facilities, ORAUT-OTIB-0004, Rev. 03 PC-2, Oak 

Ridge, Tennessee, December 6.

ORAUT (Oak Ridge Associated Universities Team), 2007, Site Profile and Technical Basis Document 

Development, ORAUT-PROC-0031, Rev. 02, Oak Ridge, Tennessee, August 17.

ORAUT (Oak Ridge Associated Universities Team), 2009, Y-12 National Security Complex –

Occupational External Dosimetry, ORAUT-TKBS-0014-6, Rev. 02, Oak Ridge, Tennessee, 

December 18.

ORAUT (Oak Ridge Associated Universities Team), 2010, Hanford Site – Occupational External 

Dose, ORAUT-TKBS-0006-6, Rev. 04, Oak Ridge, Tennessee, January 7.

Patterson, H. W., and R. H. Thomas, editors, 1994, The History of Accelerator Radiological 

Protection, Personal and Professional Memoirs, Nuclear Technology Publishing, Kent, 

England. [SRDB Ref ID: 33209]

Preisig, J. R., 1995, “Alternating Gradient Synchrotron (AGS) Bonner Multisphere Spectrometry and 

Carbon-11 Detector Results,” memorandum to E. Lessard and H. Kahnhauser, Associated 

Universities, Upton, New York, July 24. [SRDB Ref ID: 22845]

Preisig, J. R., 1997, “1997 Bonner Multisphere Spectrometer Measurements at the NSLS (April/May, 

1997), Associated Universities, Upton, New York, May 21. [SRDB Ref ID: 22843]

Reciniello, R. N., 2006, “Un-Monitored Workers,” e-mail to L. Faust, Brookhaven Science Associates, 

Brookhaven National Laboratory, April 4. [SRDB Ref ID: 33175]

Sanna, R. and O’Brien, K., 1963, “The Use of Nuclear Emulsions to Estimate Stray Neutron Spectra 

near Particle Accelerators,” Health Physics, volume 9, January, pp. 25-29.

Scarpitta, S. C., 2001, Statement of Work for In-Vitro Radiobioassay Tritium Analyses Performed at 

Brookhaven National Laboratory, BNL RCD/ASL 2001-Rev 0, Brookhaven Science 
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 103 of 115

Associates, Brookhaven National Laboratory, Upton, New York, January. [SRDB Ref ID: 

22482, p. 3]

Schroeder, G. L., D. E. Paquette, J. R. Naidu, R. J. Lee, and S. L. K. Briggs, 1998, Site Environmental 

Report for Calendar Year 1996, BNL-52543, Associated Universities, Brookhaven National 

Laboratory, Upton, New York, January. [SRDB Ref ID: 6467]

Sebelius, K., 2009, HHS Designation of Additional Members of the Special Exposure Cohort under 

the Energy Employees Occupational Illness Compensation Program Act of 2000, Designating 

a Class of Employees from Brookhaven National Laboratory, Upton, New York,” U.S. 

Department of Health and Human Services, Office of the Secretary, Washington, D.C.,

December 10. [SRDB Ref ID: 79736]

Sengupta, S., 2000, External Dosimetry Technical Basis Document, PM-PD-003, Rev. 1, Brookhaven 

Science Associates, Brookhaven National Laboratory, Upton, New York, March 31. [SRDB 

Ref ID: 33173]

Schopfer, C., 1986, “FY87 Film Badge Exchange Schedule,” memorandum to Distribution, Associated 

Universities, Brookhaven National Laboratory, Upton, New York, December 29. [SRDB Ref 

ID: 53335]

Schopfer, C., 1988, “Quality Control (QC) Test Badge Response to Calibration Sources,” 

memorandum to A. Kuehner, Associated Universities, Brookhaven National Laboratory, Upton, 

New York, May 24. [SRDB Ref ID: 55069]

Steimers, J., 1965, untitled letter to G. Welford (U.S. Atomic Energy Commission, New York 

Operations Office), Associated Universities, Brookhaven National Laboratory, Upton New 

York, September 24. [SRDB Ref ID: 6276, p. 6]

Strom, D. J., 1982, Trip Report: Argonne National Laboratory, 11/14-11/17/82, with attachments, 

University of North Carolina at Chapel Hill. North Carolina, November. [SRDB Ref ID: 12632]

Stubbings, G. 1962, “Monthly BNL Surveys – March 28 and 29, and May 2 and 3, 1962,” 

memorandum with attachment to A. J. Breslin, U.S. Atomic Energy Commission, New York 

Operations Office, New York, New York, June 1. [SRDB Ref ID: 6517, p. 74]

Sun, C., R. Reciniello, and S. Sengupta, 2004, Technical Basis Document of Brookhaven National 

Laboratory’s Internal Dosimetry Program, Rev. 4, Brookhaven Science Associates, 

Brookhaven National Laboratory, Upton, New York, March 1. [SRDB Ref ID: 22518]

Sunderman, F. W., 1947a, “Clinical X-ray Equipment,” memorandum to P. M. Morse, Associated 

Universities, Brookhaven National Laboratory, Upton, New York, September 30. [SRDB Ref 

ID: 55278]

Sunderman, F. W., 1947b, “Health Program for Employees at the Brookhaven National Laboratory,” 

memorandum to P. M. Morse, Associated Universities, Brookhaven National Laboratory, 

Upton, New York, . [SRDB Ref ID: 55277]

Sunderman, F. W., 1947c “Chest X-Rays of Laboratory Employees,” [SRDB Ref ID 55276]

Unruh, C. M., H. V. Larson, T. M. Beetle, and A. R. Keene, 1967, The Establishment and Utilization of 

Film Dosimeter Performance Criteria, BNWL-542, Pacific Northwest Laboratory, Battelle 

Memorial Institute, Richland, Washington, September. [SRDB Ref ID: 14914]
Document No. ORAUT-TKBS-0048 Revision No. 01 Effective Date: 04/26/2010 Page 104 of 115

Wilson, R. H., 1958, “The Hanford Uranium Bio-Assay Program,” p. 87, Symposium on Occupational 

Health Experience and Practices in the Uranium Industry, Held in New York City, October 15 –

17, 1958, HASL-58, U.S. Atomic Energy Commission, New York Operations Office, New York, 

New York, September. [SRDB Ref ID: 7886]

Yoder, R. C., 2005, “Landauer Film Info,” email to L. Faust, Glenwood, Illinois, December 7. [SRDB 

Ref ID: 77874]

Zukas, G., 1983, “X-ray Entrance Doses,” memorandum to W. Lehman, Associated Universities, 

Brookhaven National Laboratory, Upton, New York, October 21. [SRDB Ref ID: 55042]

Zukas, G., 1987, “X-ray Entrance Doses,” memorandum to W. Lehman, Associated Universities, 

Brookhaven National Laboratory, Upton, New York, April 21. [SRDB Ref ID: 50500]
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GLOSSARY

absorbed dose

Amount of energy (ergs or joules) deposited in a substance by ionizing radiation per unit mass 

(grams or kilograms) of the substance and measured in units of rads or grays See dose.

absorption type 

Categories for materials according to their rate of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively in the 

respiratory tract (slow solubilization). Also called solubility type. See inhalation class.

accelerator

See particle accelerator.

accreditation

For external dosimetry, the assessment of whether or not a personnel dosimetry system meets 

specific criteria. The assessment includes dosimeter performance and the associated quality 

assurance and calibration programs. 

accuracy

The characteristics of an analysis or determination that ensures that both the bias and 

precision of the resultant quantity will remain within the specified limits.

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. 

alpha particle (α) 

See alpha radiation.

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. 

ampere (A) 

International System unit of electrical current equal to 1 coulomb per second.

anterior–posterior (AP) 

Physical orientation of the body relative to a penetrating directional radiation such that the 

radiation passes through the body from the front to the back. See exposure geometry. 

backscatter

Reflection or refraction of radiation at angles over 90 degrees from its original direction.

beam quality 

Empirical measure of the ability of a polyenergetic X-ray beam to penetrate matter affected by 

the kilovoltage, anode material, voltage waveform, and filtration of an X-ray tube. The half-
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value layer in millimeters of aluminum is a typical measure of X-ray beam quality for the 

energy range used in radiography. Also called beam hardness. See filtration.

becquerel (Bq) 

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

37 billion (3.7 × 1010) Bq. 

beta particle (β) 

See beta radiation.

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.

bioassay 

Measurement of amount or concentration of radionuclide material in the body (in vivo 

measurement) or in biological material excreted or removed from the body (in vitro 

measurement) and analyzed for purposes of estimating the quantity of radioactive material in 

the body.

cladding

The outer layer of metal that encases a reactor fuel element or fissile material of the pit of a 

nuclear weapon, often made with aluminum or zirconium. In a reactor, cladding promotes the 

transfer of heat from the fuel to the coolant, and it builds up fission and activation products 

over time from the fission of the fuel.

Columbia Resin Number 39 (CR-39)

Radiosensitive material used in track-etch neutron dosimeters.

contamination 

Radioactive material in an undesired location including air, soil, buildings, animals, and 

persons.

core

Central region of a nuclear reactor where fission of the fuel takes place. 

Cosmotron (Synchrotron) 

Accelerator in which charged particles (electrons, protons, ions, etc.) are accelerated in a 

circular path achieving very high energies.

criticality

State of a radioactive mass (e.g. the core of a nuclear reactor) when the fission reaction 

becomes self-sustaining. Nuclear reactors go critical when started.

curie (Ci) 

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

cyclotron

Particle accelerator capable of large beam currents where the beam is injected in the center of 

a circular magnet. A fixed radio frequency signal applied to two D-shaped electrodes 
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accelerates the beam as it passes from one electrode to the other as the potential alternates. 

The radius of the beam increases as the energy increases.

decommissioning 

Removal of a facility from service, usually involving decontamination of radioactivity to 

specified levels and often involving demolition of the facility.

decontamination 

Reduction or removal of radioactive material from a structure, area, object, or person. 

Decontamination can occur through (1) treating the surface to remove or decrease the 

contamination or (2) allowing natural radioactive decay to occur over a period of time.

deep dose equivalent [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.

disintegrations per minute (dpm, d/m) 

Measure of radioactivity equal to the number of nuclear disintegrations in a mass per minute; 

1 dpm equals 1/60 becquerel.

DOE Laboratory Accreditation Program (DOELAP) 

Program for accreditation by DOE of DOE site personnel dosimetry and radiobioassay 

programs based on performance testing and the evaluation of associated quality assurance, 

records, and calibration programs. 

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. 

dose equivalent (H)

In units of rem or sievert, product of absorbed dose 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.

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. 

electron

Basic atomic particle with negative charge and a mass 1/1,837 that of a proton. Electrons 

surround the positively charged nucleus of the atom.
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electron-volt (eV) 

Unit equal to the energy of one electron moving through a potential difference of 1 volt 

(1.602 × 10-19 joules). The common units in nuclear physics and radiology are kiloelectronvolts (thousands) and megaelectron-volts (millions).

element 

One of the known chemical substances in which the atoms have the same number of protons. 

Elements cannot be broken down further without changing their chemical properties. 

Chemical symbols for the elements consist of either a single letter or a combination of letters, 

some of which descend from the Latin names [e.g., Au from aurum (gold), Fe from ferrum

(iron)]. This glossary indicates elements by their names. Specific isotopes appear as their 

standard chemical symbols with the number of protons and neutrons in the nucleus. 

Energy Employees Occupational Illness Compensation Program Act of 2000, as 

amended (EEOICPA; 42 U.S.C. § 7384 et seq.) 

Law that provides for evaluation of cause and potential compensation for energy employees 

who have certain types of cancer.

entrance skin exposure (ESE) 

Air kerma without backscatter at the point of entry into the body. Also called entrance kerma. 

See kerma.

exchange period (frequency)

Period (weekly, biweekly, monthly, quarterly, etc.) for routine exchange of dosimeters.

exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

exposure geometry 

Orientation (physical positioning) of a person or object in relation to a radiation source. This 

geometry is a factor in the radiation dose to various parts of the body. See anterior–posterior, 

posterior–anterior, and lateral in relation to radiology.

favorable to claimants

In relation to dose reconstruction for probability of causation analysis, having the property of 

ensuring that there is no underestimation of potential dose, which often means the assumption 

of a value that indicates a higher dose than is likely to have actually occurred in the absence of 

more accurate information. See probability of causation.

film

(1) In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. See film dosimeter. (2) X-ray film.

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.
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filter

Material used in a dosimeter to adjust radiation response to provide an improved tissue 

equivalent or dose response.

filtration 

The process of filtering an X-ray beam, usually with millimeter thicknesses of aluminum 

material between the X-ray source and the film that preferentially absorbs photons from the 

beam. Usually measured in equivalent millimeters of aluminum. See beam quality and halfvalue layer.

fission

Splitting of the nucleus of an atom (usually of a heavy element) into at least two other nuclei 

and the release of a relatively large amount of energy. This transformation usually releases 

two or three neutrons.

fission product

Radionuclides produced by fission or by the subsequent radioactive decay of radionuclides. 

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 photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus. 

gamma ray

See gamma radiation.

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.

half-value layer (HVL) 

Thickness of a specified substance, usually specified in equivalent millimeters of aluminum, 

that filters an X-ray beam to reduce the kerma rate by one-half. See filtration. 

hot cell 

Shielded laboratory for handling of radioactive materials with the aid of remotely operated 

manipulators. The walls and windows are made of materials that protect workers from 

radiation. 

in vitro bioassay

Measurements to determine the presence of or to estimate the amount of radioactive material 

in the excreta or in other biological materials removed from the body.

in vivo bioassay

The measurements of radioactive material in the human body utilizing instrumentation that 

detects radiation emitted from the radioactive material in the body. 

inhalation class 

Former respiratory tract inhalation classification scheme developed by the International 

Council on Radiological Protection for inhaled material according to its rate of clearance from 

the pulmonary region of the lung. Materials were classified as D (days, half-life less than 10 
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days), W (weeks, 10 to 100 days), or Y (years, more than 100 days). See absorption type, 

which superseded this concept. 

internal dose 

Dose received from radioactive material in the body.

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. See 

alpha radiation, beta radiation, gamma radiation, photon radiation, and X-ray radiation.

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties. See element.

kerma

Measure in units of absorbed dose (usually grays but sometimes rads) of the energy released 

by radiation from a given amount of a substance. Kerma is the sum of the initial kinetic 

energies of all the charged ionizing particles liberated by uncharged ionizing particles 

(neutrons and photons) per unit mass of a specified material. Free-in-air kerma refers to the 

amount of radiation at a location before adjustment for any external shielding from structures 

or terrain. The word derives from kinetic energy released per unit mass.

lateral (LAT) 

Orientation of the body during an X-ray procedure in which the X-rays pass from one side of 

the body to the other. See exposure geometry.

linear accelerator (LINAC) 

Straight single-pass particle accelerator in which radio frequencies accelerate the beam over 

the length of the accelerator.

lumbar spine

Region of the spine including the five lumbar vertebrae (lower back).

megaelectron-volt (MeV) 

Unit of particle energy equal to 1 million (1 × 106

) electron-volts. 

milliampere-second (mAs) 

In relation to radiography, product of the average X-ray beam current in milliamperes and the 

time of the exposure in seconds. These factors are selectable on the control panel of most 

medical X-ray equipment.

natural uranium

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 becquerel per kilogram (0.7 picocuries 

per gram).
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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.

neutron radiation 

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons. See neutron.

nonpenetrating dose (NP, NPEN)

Dose from beta and lower energy photon (X-ray and gamma) radiation which does not 

penetrate the skin. It is often determined from the open window dose minus the shielded 

window dose. See dose.

nuclear emulsion 

Thick photographic coating in which the tracks of various fundamental particles show as black 

traces after development. The number of tracks in a given area is a measure of the dose from 

that radiation. See nuclear track emulsion, type A. 

nuclear energy 

Energy released by nuclear reaction, some of which can be ionizing radiation. Of particular 

importance is the energy released when a neutron initiates fission or when two nuclei join 

together under millions of degrees of heat (fusion). Also called atomic energy.

nuclear track emulsion, type A (NTA) 

Film sensitive to fast neutrons made by the Eastman Kodak. The developed image has tracks 

caused by neutrons that become visible under oil immersion with about 1,000-power 

magnification.

nuclide 

Stable or unstable isotope of any element. Nuclide relates to the atomic mass, which is the 

sum of the number of protons and neutrons in the nucleus of an atom. A radionuclide is an 

unstable nuclide.

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, but does include dose from occupationally required radiographic examinations that 

were part of medical screening.

occupational environmental dose

Dose received while on the grounds of a site but not inside a building or other facility.

occupational medical dose 

Dose from X-ray procedures performed for medical screening of workers as part of an 

occupational health program. Doses from X-rays used to diagnose diseases or injuries, even 

if incurred on the job, are not considered occupational and are therefore not eligible to be 

included in dose reconstruction under EEOICPA.
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open window (OW)

Area of a film dosimeter that has little to no radiation shielding (e.g., only a holder and visible 

light protection). See film dosimeter.

particle accelerator 

Device that accelerates ions using magnetic and/or electrostatic fields for focusing and 

redirecting ion beams. The main purposes of accelerators are the investigation of high-energy 

particle behavior and production of synthetic isotopes.

penetrating dose (PEN) 

Dose from higher energy photon (gamma and X-ray) radiation that penetrates the outer layers 

of the skin. See dose.

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. 

phantom

Any structure that contains one or more tissue substitutes (any material that simulates a body 

of tissue in its interaction with ionizing radiation) and is used to simulate radiation interactions 

in the human body. Phantoms are primarily used in the calibration of in vivo counters and 

dosimeters.

photofluorography (PFG) 

Historical radiographic technique used for chest images for screening a large number of 

people in a short period of time. The X-ray image produced on a fluorescent screen was 

photographed on 4- by 5-inch film. PFG was the primary method of screening large 

populations for tuberculosis before the advent of nonradiographic screening methods. Also 

called fluorography or mass miniature radiography. Not to be confused with fluoroscopy.

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz. 

photon radiation 

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays. 

PM

A procedure detailing specific actions or directions and usually limited to one service or 

activity.

posterior-anterior (PA) 

Physical orientation of the body relative to a penetrating directional radiation field such that the 

radiation passes through the body from the back to the front. See exposure geometry.
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probability of causation

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Act, the percent likelihood, at the 99th percentile, that a worker incurred a 

particular cancer from occupational exposure to radiation.

proton 

Basic nuclear particle with a positive electrical charge and mass slightly less than that of a 

neutron. There are protons in the nuclei of every atom, and the number of protons is the 

atomic number, which determines the chemical element. See element.

quality factor (QF) 

Principal modifying factor (which depends on the collision stopping power for charged 

particles) that is employed to derive dose equivalent from absorbed dose. The quality factor 

multiplied by the absorbed dose yields the dose equivalent. See dose, relative biological 

effectiveness, and weighting factor.

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 joules 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. 

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

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei. See radionuclide. 

radiograph 

Image produced on film by gamma rays or X-rays. See radiology.

radiography 

The process of producing images on film (or other media) with radiation. 

radiology 

Medical science and specialty of producing images on radiographic film or other media, which 

are used to identify, diagnose, and or treat diseases, injuries, or other conditions. 

radionuclide

Radioactive nuclide. See radioactive and nuclide. 

reactor 

Device in which a fission chain reaction occurs under controlled conditions to produce heat or 

useful radiation for experimental purposes or to generate electrical power or nuclear fuel. 

reactor elements 

Fabricated fuel and target components inserted into reactors.
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relative biological effectiveness (RBE) 

Ratio of the absorbed dose of a reference radiation to the absorbed dose of a test radiation 

that produces the same biological effects, other conditions being equal. A factor applied to 

account for differences between the amount of cancer effect produced by different forms of 

radiation.

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 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.

rep

Historical quantity of radiation (usually other than X-ray or gamma radiation) originally defined 

as 83 ergs absorbed per gram in the body and redefined in the 1940s or early 1950s as the 

amount that would liberate the same amount of energy (93 ergs per gram) as 1 roentgen of Xor gamma rays. Replaced by the gray in the International System of Units; 1 rep is 

approximately equal to 8.38 milligray. The word derives from roentgen equivalent physical.

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 coulombs 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). 

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.

skin dose

See shallow dose equivalent.

technique 

Combination of X-ray machine settings used to produce radiographs, which consists of the 

kilovoltage, tube current (milliamperes), and exposure time (seconds). The last two 

parameters are often multiplied to yield the electric charge that has crossed the X-ray tube 

during the exposure in units of milliampere-seconds. Any combination of time and tube 

current that produces a given product in milliampere-seconds produces the same exposure for 

a fixed peak kilovoltage.

source-to-image distance (SID) 

Distance from the X-ray machine target (anode) to the plane of the image receptor (film). This 

distance is standardized for typical radiographic procedures. Chest X-rays, for example, are 

performed at a 72-inch SID.
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tandem Van de Graaff accelerator 

Accelerator in which charge exchange (negative to positive) occurs in the terminal on either a 

thin foil or in a gas stripper tube.

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. 

tissue equivalent

Substance with response to radiation equivalent to tissue. A tissue-equivalent response is an 

important consideration in the design and fabrication of radiation measuring instruments and 

dosimeters.

type

See absorption type.

U.S. Atomic Energy Commission (AEC) 

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The U.S. Energy Research and Development 

Administration and the U.S. Nuclear Regulatory Commission assumed separate duties from 

the AEC in 1974. The U.S. Department of Energy succeeded the U.S. Energy Research and 

Development Administration in 1979.

weighting factor

The ratio of the stochastic risk arising from tissue T to the total risk when the whole body is 

irradiated uniformly. (ICRP 1978)

whole-body counter (WBC) 

Equipment used to perform in vivo bioassay. Radiation emitted from radioactive material 

deposited throughout the body is measured. 

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 

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom. 

zirconium

Metallic element with atomic number 40. Zirconium is highly resistant to corrosion, and it is 

alloyed with aluminum to make cladding for nuclear fuel and sometimes in small amounts with 

the fuel itself.