ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 131

DOE Review Release 02/14/2013

Document Title:

Site Profile for the Brookhaven National 

Laboratory

Document Number: ORAUT-TKBS-0048

Revision: 02

Effective Date: 02/07/2013

Type of Document: TBD

Supersedes: Revision 01 

Subject Expert(s): Lori Arent, Samuel Chu, Leo Faust, Robert L. Morris, Eugene Potter, and 

Elyse M. Thomas

Approval: Signature on File Approval Date: 01/28/2013

Dennis L. Strenge, Document Owner

Concurrence: Signature on File Concurrence Date: 01/28/2013

John M. Byrne, Objective 1 Manager

Concurrence: Keith A. McCartney Signature on File for Concurrence Date: 01/28/2013

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 01/28/2013

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 02/07/2013

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. 02 Effective Date: 02/07/2013 Page 2 of 131

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.

02/07/2013 02 Revision initiated to incorporate extension of SEC period through 

1993, SC&A comments and DCAS concerns. Section 1 was revised 

to include a new Section 1.2 to address SEC issues. Section 2 was 

revised to include updated information on facilities including 

expansion of information on the LMFR. Section 3 was revised to 

correct PFG dose values (Section 3.7) and set medical X-ray 

uncertainty to 30% (Section 3.8). Section 4.4 was revised to update 

external ambient dose values (Table 4-1) and extend ambient intakes 

to 2010 (Tables 4-2 and 4-3). Section 5 was updated to provide 

additional information on minimum detection limits for bioassay 

sample analyses (Sections 5.4 and 5.5) and guidance on evaluation 

of internal dose to volunteers participating in human testing by the 

Medical Department (new Section 5.4.5). Section 6 was revised to 

update discussion of neutron dosimetry (Section 6.4), add annual 

average worker dose for unmonitored workers (new Section 6.5.2), 

expand guidance on neutron track fading and angular dependence of 

NTA film (Section 6.10), add guidance on evaluation of external dose 

to volunteers participating in human testing by the Medical 

Department (new Section 6.11), and revise uncertainty factor 

guidance (Section 6.12). Incorporates formal internal and NIOSH 

review comments. Information added on three facilities in response 

to SC&A issues and findings in Sections 2.4.16.3, 2.4.16.4, 2.4.16.5, 

5.4 and Tables 2-1, 6-4, and 3-1. Constitutes a total rewrite of the 

document. Training required: As determined by the Objective 

Manager. Initiated by Dennis L. Strenge.
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 3 of 131

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 7

1.0 Introduction ............................................................................................................................ 11

1.1 Purpose...................................................................................................................... 12

1.2 Special Exposure Cohort Petition Information............................................................. 12

1.3 Scope......................................................................................................................... 13

1.3.1 Site Description ............................................................................................. 13

1.3.2 Occupational Medical Dose ........................................................................... 13

1.3.3 Occupational Environmental Dose................................................................. 13

1.3.4 Occupational Internal Dose ........................................................................... 13

1.3.5 Occupational External Dosimetry................................................................... 14

1.3.6 Attributions and Annotations.......................................................................... 14

2.0 Site Description...................................................................................................................... 14

2.1 Introduction................................................................................................................. 14

2.2 Purpose...................................................................................................................... 14

2.3 Scope......................................................................................................................... 14

2.4 Site Description and General Information ................................................................... 14

2.4.1 Brookhaven Graphite Research Reactor, Buildings 701, 704, 708, and 

709 ................................................................................................................ 26

2.4.2 High Flux Beam Reactor Complex, Buildings 750, 705, and 751................... 27

2.4.3 Medical Research Center, Building 490......................................................... 27

2.4.4 Brookhaven Medical Research Reactor, Building 491 ................................... 28

2.4.5 Positron Emission Tomography, Building 906................................................ 28

2.4.6 Cosmotron, Building 902 ............................................................................... 28

2.4.7 Alternating Gradient Synchrotron, 900 Series Buildings................................. 29

2.4.8 Brookhaven LINAC Isotope Producer, Building 931....................................... 33

2.4.9 Tandem Van de Graaff, Building 901............................................................. 33

2.4.10 National Synchrotron Light Source, Building 725........................................... 35

2.4.11 Relativistic Heavy Ion Collider Complex, Building 1000 Series ...................... 35

2.4.12 Hot Laboratory, Building 801 ......................................................................... 38

2.4.13 High Intensity Radiation Development Laboratory, Building 830.................... 39

2.4.14 Chemistry Department, Building 555 ............................................................. 39

2.4.15 Physics Department, Building 510 ................................................................. 40

2.4.16 Other Miscellaneous Radiological Research Facilities................................... 41

2.4.16.1 Liquid Metal Fuel Reactor Support, Building 820 ........................... 41

2.4.16.2 Hot Machine Shop, Building 530.................................................... 41

2.4.16.3 Health Physics and Safety Instrumentation, Buildings 

535 and 356................................................................................... 41

2.4.16.4 Biology Department, Building 463.................................................. 42

2.4.16.5 Radiation Calibrations Facility, Building 348 .................................. 42

2.4.17 Waste Management Facility, Radwaste Reclamation Building, 

Building 865................................................................................................... 42

2.4.18 Radiological Waste Decontamination Facility, Building 650 ........................... 43

2.5 Isotopes of Concern ................................................................................................... 43

2.6 Magnitude of Site Activity ........................................................................................... 43

2.7 Major Incidents ........................................................................................................... 44

2.8 Radiological Access Controls ..................................................................................... 44
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3.0 Occupational Medical Dose.................................................................................................... 45

3.1 Purpose...................................................................................................................... 45

3.2 Scope......................................................................................................................... 45

3.3 Examination Frequency.............................................................................................. 46

3.4 Equipment and Techniques ........................................................................................ 47

3.5 Incident Air Kerma...................................................................................................... 48

3.6 Organ Dose Calculations............................................................................................ 50

3.7 Dose Reconstruction .................................................................................................. 51

3.8 Uncertainty ................................................................................................................. 55

4.0 Occupational Environmental Dose ......................................................................................... 55

4.1 Introduction................................................................................................................. 55

4.2 Purpose...................................................................................................................... 56

4.3 Scope......................................................................................................................... 56

4.4 Ambient External Radiation ........................................................................................ 56

4.4.1 Onsite Monitoring .......................................................................................... 56

4.4.2 Applicability for Dose Reconstruction............................................................. 59

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

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

4.5.2 Estimates of Potential Inhalation Intake......................................................... 60

4.6 Ingestion..................................................................................................................... 64

4.7 Uncertainty ................................................................................................................. 64

5.0 Occupational Internal Dose.................................................................................................... 65

5.1 Introduction................................................................................................................. 65

5.2 Purpose...................................................................................................................... 65

5.3 Scope......................................................................................................................... 65

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

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

5.4.2 Tritium Exposure Evaluation Form................................................................. 66

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

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

5.4.5 Voluntary Intakes of Radionuclides................................................................ 67

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

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

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

5.5.2 Fecal Sample Analysis .................................................................................. 69

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

Practices .................................................................................................................... 69

5.6.1 Whole-Body Counting.................................................................................... 69

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

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

5.8 Detection Limits.......................................................................................................... 71

5.9 Excreta Sample Kit Codes.......................................................................................... 74

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

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

5.12 Workplace Monitoring Data ........................................................................................ 78

5.13 Radon......................................................................................................................... 78

5.14 Guidance for Dose Reconstructors............................................................................. 79

6.0 Occupational External Dosimetry ........................................................................................... 79

6.1 Introduction................................................................................................................. 79

6.2 Purpose...................................................................................................................... 79
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6.3 Scope......................................................................................................................... 79

6.4 BNL Dosimetry Program............................................................................................. 79

6.4.1 Early Dosimetry for High-Energy Proton Accelerators ................................... 80

6.4.2 Neutron Dosimetry......................................................................................... 81

6.4.3 Special Dosimetry.......................................................................................... 82

6.4.4 Lost or Destroyed Dosimeters ....................................................................... 82

6.5 Missed Dose............................................................................................................... 82

6.5.1 Unmonitored Worker ..................................................................................... 82

6.5.2 Annual Average Worker Dose ....................................................................... 83

6.6 Radiation Energies and Percentages at Selected BNL Facilities ................................ 84

6.7 Neutron Radiations and Percentages ......................................................................... 87

6.8 Recorded Dose Practices ........................................................................................... 89

6.9 Interpretation of Reported Doses................................................................................ 90

6.10 Adjustments to Recorded Dose .................................................................................. 91

6.10.1 NTA Track Fading ......................................................................................... 91

6.10.2 NTA Angular Dependence............................................................................. 91

6.11 Voluntary Neutron Radiation Experiments .................................................................. 93

6.12 Bias and Uncertainty .................................................................................................. 94

7.0 Attributions and Annotations .................................................................................................. 96

References ........................................................................................................................................ 98

Glossary .......................................................................................................................................... 117

ATTACHMENT A, TECHNICAL BASIS FOR THE MINIMUM DETECTABLE DOSE 

FOR NEUTRON NTA FILM AT BNL .................................................................. 129
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LIST OF TABLES

TABLE TITLE PAGE

2-1 Site information of major nuclear research facilities at BNL.................................................... 18

2-2 Major incidents.......................................................................................................................44

3-1 Default frequency of chest X-rays at BNL............................................................................... 46

3-2 X-ray equipment used at BNL and incident air kerma values ................................................. 50

3-3 Organ doses for PFG, PA, and LAT 14- by 17-in. chest radiography ..................................... 51

3-4 Skin dose guidance for various chest projections and periods ............................................... 52

3-5 Skin dose from various chest projections ............................................................................... 53

4-1 External gamma radiation dose.............................................................................................. 56

4-2 Maximum annual environmental occupational radionuclide inhalation.................................... 61

4-3 Maximum annual environmental occupational radionuclide inhalation, 2005 to 2010 ............. 64

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

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

5-3 Detection limits for in vitro bioassay ....................................................................................... 71

5-4 Detection limits for in vivo bioassay........................................................................................ 73

5-5 Solubility type and fraction activity by facility .......................................................................... 75

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

6-1 Estimated annual missed photon, beta, and neutron dose ..................................................... 82

6-2 Annual average workplace external dose............................................................................... 84

6-3 Electron and photon energies and percentages for reactors and accelerators ....................... 86

6-4 Electron and photon energies and percentages (for support facilities) ................................... 86

6-5 Facilities, neutron energies, percentages, and correction factors ........................................... 87

6-6 Recorded dose practices........................................................................................................ 90

6-7 Interpretation of reported data................................................................................................ 90

6-8 Summary of NTA fading studies at BNL................................................................................. 91

6-9 Ratio of NTA film neutron dose for rotational geometry with angular correction to AP 

geometry................................................................................................................................92

6-10 Landauer gamma film, NTA film, and CR-39 specifications.................................................... 95

6-11 Uncertainty factors .................................................................................................................96

A-1 Data from a 1955 background study and the Poisson fit to the data..................................... 129

A-2 1974 calibration factors applied to the calculated LC in tracks .............................................. 131

LIST OF FIGURES

FIGURE TITLE PAGE

2-1 Aerial view of Brookhaven National Laboratory ...................................................................... 15

2-2 Nuclear and radioactivity sites at BNL.................................................................................... 17

2-3 The AGS complex..................................................................................................................30

2-4 The Relativistic Heavy Ion Collider complex........................................................................... 36

2-5 RHIC ring intersection points.................................................................................................. 37

6-1 Neutron spectrum in the thermal column with the converter plate and in the 

instrument tunnel ...................................................................................................................88

6-2 Neutron energy spectrum at the Cosmotron........................................................................... 89

6-3 Logic diagram for application of NTA film angular correction factor........................................ 93

A-1 Plot of observed and expected tracks for background control films ...................................... 130
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AGS Alternating Gradient Synchrotron

AMAD activity median aerodynamic diameter

AP anterior-posterior

ATR AGS-To-RHIC

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

BRAHMS Broad Range Hadron Magnetic Spectrometers

CERN European Organization for Nuclear Research (formerly Conseil Européen pour la 

Recherche Nucléaire)

CFR Code of Federal Regulations

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

DØ D-zero

DCF dose conversion factor

DTL Drift-Tube-LINAC

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

DOL U.S. Department of Labor

dpm disintegrations per minute

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EML Environmental Measurement Laboratory

ENSD entrance skin dose

ESE entrance skin exposure

EXSD exit skin dose

F fast absorption type

FOV field of view

ft foot

g gram

GeV gigaelectron-volt, 1 billion electron-volts

GIF Gamma Irradiation Facility

GSD geometric standard deviation

HASL Health and Safety Laboratory

HEBT High-Energy Beam Transport
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HEPA high-efficiency particulate air (filter)

HFBR High Flux Beam Reactor

HIRDL High Intensity Radiation Development Laboratory

hr hour

HVL half-value layer

HWMF Hazardous Waste Management Facility

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

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

LET linear energy transfer

LINAC linear accelerator

LMFR Liquid Metal Fuel Reactor

M moderate absorption type

m meter

mA milliampere

mAs milliampere-second

mCi millicurie

MDA minimum detectable activity or amount

MDD minimum detectable dose

MDL minimum detectable level

MEL Metallurgical Evaluation Laboratory

MeV megaelectron-volt, 1 million electron-volts

MFP mixed fission product

mL milliliter

mm millimeter

mR milliroentgen

mrad millirad

MRC Medical Research Center

mrem millirem

mrep millirep

MW megawatt

n neutron

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 National Aeronautics and Space Administration Space Radiation Laboratory

NTA nuclear track emulsion, type A

NVLAP National Voluntary Laboratory Accreditation Program
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NYOO New York Operations Office

OEF Office of Environmental Restoration

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

OW open window

p proton

PA posterior-anterior

pCi picocurie

PET positron emission tomography

PFG photofluorography

PHENIX Pioneering High Energy Nuclear Interaction Experiment

POC probability of causation

ppm parts per million

QF quality factor

R roentgen

R&D research and development

radwaste radioactive waste

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 absorption type

s second

S&EP Safety and Environmental Protection (Division)

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

STAR Solenoidal Tracker (Detector)

STRL Severn-Trent Richland Laboratory

t tritium

T tritium

TARL TestAmerica Richland Laboratory

TLD thermoluminescent dosimeter

TPL Target Processing Laboratory

TRISTAN Terrific Reactor Isotope Separator to Analyze Nuclides

U.S.C. United States Code

VUV vacuum ultraviolet

WB whole body

WBC whole-body counter

WCF Waste Concentration Facility

wk week

WMF Waste Management Facility
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yr year

μA microampere

μCi microcurie

μg microgram

μm micrometer

§ section or sections
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1.0 INTRODUCTION

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

] 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) restrict the “performance of duty”

referred to in 42 U S. C. § 7384n(b) to nuclear weapons work (NIOSH 2010).

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 2010):

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

conventional structures

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

reasons

 1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
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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). Section 1.2 details the BNL classes in the Special Exposure 

Cohort (SEC), and Section 1.3 describes the document content.

1.2 SPECIAL EXPOSURE COHORT PETITION INFORMATION

Classes Added to the SEC

Brookhaven National Laboratory (1947 to 1979)

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

contractors and subcontractors who worked in any area at Brookhaven National 

Laboratory in Upton, New York, from January 1, 1947, through December 31, 1979, for 

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

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

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

Cohort.

Based on the findings and recommendations of NIOSH and the Advisory Board on Radiation and 

Worker Health, the Secretary of the U.S. Department 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, about 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).

Brookhaven National Laboratory (1980 to 1993)

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

contractors and subcontractors who worked in any area at Brookhaven National 

Laboratory in Upton, New York, from January 1, 1980 through December 31, 1993, for 

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

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

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

Cohort.

Through the course of ongoing dose reconstruction and research, NIOSH has determined that the 

issues identified in the first Brookhaven National Laboratory evaluation (NIOSH 2009) carry over into 

later periods at the site. Specifically, NIOSH has determined that, due to ongoing difficulty of 

consistently obtaining all requested personnel monitoring records for individual BNL claims, NIOSH is 

unable to estimate with sufficient accuracy internal exposures for workers at Brookhaven National 

Laboratory during the period from January 1, 1980, through December 31, 1993. NIOSH has found 

that there is a potential for undocumented worker movements across the site, and that there is limited 

claimant-specific information about work locations. NIOSH has found that a determination cannot 

always be made as to whether or not an employee worked in technical areas with a history of 

radioactive material use, or whether an employee should have been monitored for radiological 

exposures. Considering this information, NIOSH is unable to eliminate any specific worker from 

potential exposure scenarios based on assigned work location (Sebelius 2012).

Although NIOSH has found that it is not possible to completely reconstruct radiation doses for the 

SEC classes discussed above, it intends to use internal and external monitoring data that might 

become available for an individual claim (and that can be interpreted using existing NIOSH dose 

reconstruction processes or procedures). Therefore, dose reconstructions for individuals employed at 
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the Brookhaven National Laboratory facility during the SEC periods stated above, but who do not 

qualify for inclusion in the SEC, can be performed using these data as appropriate to support a partial 

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

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

the SEC periods.

1.3 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.

1.3.1 Site Description

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.

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 (PET) facility, and the medical research reactor; and the 

operation of the accelerators and 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.

1.3.2 Occupational Medical Dose

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

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

scientists employed for more than 3 months, were given an entry and annual physical examination 

including a posterior-anterior (PA) chest X-ray. Later, the annual frequency might 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.

1.3.3 Occupational Environmental Dose

Section 4.0 discusses the radiation doses received by workers at BNL while on the site but outside 

the facilities. From 1967 to 1984, ambient gamma radiation dose rates are available from four 

perimeter stations on the northwest, southwest, southeast, and northeast. Beginning in 1985, the 

onsite monitoring program was expanded to approximately 20 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 wellshielded radiation source. The radionuclides emitted from the BNL stacks are identified and their 

annual inhalation in becquerels per year are listed. Neither potable groundwater nor soil ingestion 

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

1.3.4 Occupational Internal Dose

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

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

with worker radiation exposures that are covered by EEOICPA. It lists the types of routine monitoring 
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that constitute the in vitro and the in vivo dose control programs as well as the minimum detectable 

activities (MDAs), counting methods, and reporting practices. The section addresses the uncertainty 

for BNL exposure and dose records.

1.3.5 Occupational External Dosimetry

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 that are associated with the monitoring of 

external dose. The missed doses were basically those from dosimeter minimum detectable levels 

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

particles that were generated by the accelerators. Neutron doses were measured with nuclear track 

emulsion, type A (NTA) film until 1995, after which Columbia Resin Number 39 (CR-39) has been 

used.

1.3.6 Attributions and Annotations

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

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

2.0 SITE DESCRIPTION

2.1 INTRODUCTION

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 periods of the process operations, and 

major site incidents that resulted in significant internal or external exposure to one or more persons.

2.3 SCOPE

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

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

Much of the information is taken from the book “Making Physics, A Biography of Brookhaven 

National Laboratory, 1946 – 1972” by Crease (1999).

2.4 SITE DESCRIPTION AND GENERAL INFORMATION

Figure 2-1 shows an aerial view of BNL.
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Figure 2-1. Aerial view of Brookhaven National Laboratory (BNL 2009a).

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. It 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 

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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Particle Accelerators

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

accelerator to achieve billion electron-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 

metals. These metals, which become activated by the proton beam, are then processed at the Target 

Processing Laboratory (TPL) 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 uses 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 TPL (also called the 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 

radioisotopes 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.
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Waste Management

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

Facility (HWMF) in its entirety and consolidated 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; it 

also lists radionuclides of concern (see Section 2.5). The following subsections comprise further 

descriptions of the research facilities of significance on the BNL site and provide background 

information relevant to dose reconstruction.

Figure 2-2. Nuclear and radioactivity sites at BNL [1].

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

Section Name Description Period

Building 

No. Note Radionuclide

2.4.1 BGRR U-fueled, graphite-moderated 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

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 HEPAa filter elements 1950–unknown 708 Filter elements still in 

place in 1997

Same as above

2.4.2 HFBR 30- to 60-MW thermal, heavy-water-moderated 

research reactor

1965–1999 750 Ceased operation due to 

leak in spent fuel pool in 

1997

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 Produced extremely low-energy neutrons by 

moderating neutrons with liquid hydrogen

1980–1996 751 None

2.4.3 MRC Medical Department 1959–present 490 Dispersible 

radioisotopes, PuBe 

sources (removed)

None
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Section Name Description Period

Building 

No. Note Radionuclide

2.4.4 BMRR Enriched U-fueled, light-water-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

Zn-65

Zn-69m

2.4.5 PET Imaging 

Laboratory

PET scanners for nuclear medicine and 

neurosciences research

1977–present 906 Use of positron emitters 

(C-11, F-18, N-13, and 

O-18)

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 AGS tunnel, short-lived 

activation products in 

tunnel

C-11

N-13

O-15

Long-lived activation C-14
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Section Name Description Period

Building 

No. Note Radionuclide

products Na-22, Na-24

Cr-51

Mn-54

Fe-55, Fe-59

Co-60

Ni-59, Ni-63

Zn-65

2.4.7 200-MeV LINAC Preaccelerator and LINAC providing proton 

source

1967–present 930 None

2.4.7 LINAC cooling 

support service

Provides water-cooling services for the LINAC 

complex

1967–present 930 MER Contains activated water None

2.4.7 AGS Booster 

Tunnel

Accumulator booster ring 1967–present 942 Short-lived activation 

products in tunnel

C-11

N-13

O-15

Long-lived activation 

products

C-14

Na-22, Na-24

Cr-51

Mn-54

Fe-55, Fe-59

Co-60

Ni-59, Ni-63

Zn-65

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 C-14

Na-22, Na-24

Cr-51

Mn-54

Fe-55, Fe-59

Co-60

Ni-59, Ni-63

Zn-65

2.4.7 AGS ventilation 

system 

fanhouses

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

to proximity to the 

tunnel; activated water

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 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
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Section Name Description Period

Building 

No. Note Radionuclide

2.4.7 AGS Department Office, field testing shops, machine shops, high 

bays.

1960–present 911 Storage and repair of 

activated components

C-14

Na-22, Na-24

Cr-51

Mn-54

Fe-55, Fe-59

Co-60

Ni-59, Ni-63

Zn-65

2.4.7 g-2/Bubble 

Chamber

30-in. and 80-in. Bubble Chamber 1962–1978 919 Tritium release in 1973 H-3

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

Storage of activated 

components

C-14

Na-22, Na-24

Cr-51

Mn-54

Fe-55, Fe-59

Co-60

Ni-59, Ni-63

Zn-65

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

1972–present 931B Buildings 931A and 

931C have no nuclear 

hazard

As-72

As-74

Be-7

Co-57

Co-58

Co-60

Cs-132

Cs-137

H-3

Mn-54

Na-22

O-15

Sc-46

Co-56

Ga-68

Ge-69

I-126
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Section Name Description Period

Building 

No. Note Radionuclide

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 Dynamitron (positron 

accelerator) operations 

ceased in 1989

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 Produces intense sources of X-ray, ultraviolet, 

and infrared radiation

1981–present 725

2.4.10 NSLS 

Development 

Laboratory

Houses a LINAC 1981–present 729 Has previous radioactive 

works

2.4.11 RHIC Complex High-energy particle accelerator and collider 1999–present 1002 Uranium plate machining

BRAHMS experimental hall 1981–present 1004A Has previous radioactive 

works

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
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Section Name Description Period

Building 

No. Note Radionuclide

Future experimental hall 1994–present

2.4.12 Hot Laboratory Radioisotope processing and production, TPL, 

and radiopharmaceutical research

1951–present 801 Explosion of UF6/BrF3 on 

5/15/1957; a few people 

were injured

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 GIF

1951–present

GIF 1970–

present

830 GIF contained 

approximately 35,000 Ci 

of Co-60 (removed) and 

megacurie Co-60 and 

Cs-173 sources 

(removed)

Co-60

Cs-137

2.4.14 Chemistry 

Department

Studies of chemical kinetics, hot-atom effects, 

nuclear chemical studies of nuclear decay 

schemes and nuclear reaction cross-sections, 

solar neutrinos, and radiocarbon dating

1966–present 555
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Section Name Description Period

Building 

No. Note Radionuclide

2.4.15 Physics 

Department

Laboratories, office space, a machine shop, 

and basement area

1962–present 510 Radioactive material 

storage, counting 

laboratories

2.4.16.1 LMFR Support 

Facility

R&D work for the LMFR 1957–1975 820 U-235

U-238

U-234

2.4.16.2 Hot Machine 

Shop

Hot Machine Shop 1947–1975 530

2.4.16.3 Instrument 

Division

R&D 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

2.4.16.3 Instrument 

Division, Physics 

Department

Irradiation facility for material R&D 1960 – present 356 Sealed sources Co-60

2.4.16.4 CERF Gamma irradiation facility 1951 – present 463 Sealed sources Co-60

Cs-137

2.4.16.5 S&EP/348 Radiological Calibrations Facility 1959 - present 348 Sealed beta, gamma, 

and neutron sources

Co-60

Cs-137

PuBe

Ra-226

Sr-90

U Nat

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

H-3

Mn-54

Na-22

Se-75
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Section Name Description Period

Building 

No. Note Radionuclide

Zn-65

2.4.17 Waste 

Management 

Incinerator

Low-level waste incinerator Early 1970s to 

early 1990s

444 Incinerated animal 

carcasses with no 

radiological content

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

a. HEPA = high-efficiency particulate air.
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2.4.1 Brookhaven Graphite Research Reactor, Buildings 701, 704, 708, and 709

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 

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; it 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 

elements and operated at power levels 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 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 in 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).

Building 709

Building 709 was 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 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 over 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.
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By 1972, the canal was finally drained, cleaned with detergents and water, and covered with concrete 

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

2.4.2 High Flux Beam Reactor Complex, Buildings 750, 705, and 751

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 and is 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 provided increased uranium loading in anticipation of an overall facility upgrade. In 

1982, several major modifications were completed that allowed operation at power levels up to 

60 MW.

In 1990, as a result of questions 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 the HFBR’s spent 

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

permanently close the HFBR.

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 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).

Building 751

In 1980, the Cold Neutron Facility (CNF) was installed at beamline 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.

2.4.3 Medical Research Center, Building 490

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.

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 (WBC), and two single-photon emission 
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computed tomography cameras. Some of these facilities contain or contained 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 Brookhaven Medical Research Reactor, Building 491

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

research. It 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 

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

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

reflector that surrounds 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;

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 Positron Emission Tomography, Building 906

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.

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; today it is called the Pet Imaging Laboratory (BNL 1997f, 

p. 5).

2.4.6 Cosmotron, Building 902

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.

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 (R&D) 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)
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2.4.7 Alternating Gradient Synchrotron, 900 Series Buildings

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.

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.

The AGS LINAC, 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 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 beamstop. 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 per 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 to 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 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.
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Figure 2-3. The AGS complex (BNL 1997h).

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. However, because 

the involved radionuclides 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 
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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 

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. 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.

A majority of the AGS experiments, about 400, have occurred in this large area. The AGS 

experimental floor also contains a section of the AGS ring. Surrounding the beamlines 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 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 beamlines when protons are 

being extracted to these areas. There are activated components in the beamlines, 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 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 designated 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. In addition, air handled in the fanhouses is from the 

AGS tunnel air and can 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 (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 
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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).

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. The basement 

also contains 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 beamline, 

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, from 1972 to 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, from 1979 to 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 in relation 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 

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 cool-down 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 to 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 and stands, 

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

2.4.8 Brookhaven LINAC Isotope Producer, Building 931

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

mound (BNL 1997j). 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.

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 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 forklift 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 Tandem Van de Graaff, Building 901

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.

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

home of 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-LINAC (DTL) 

area, laboratory space associated with the Department of Applied Science, 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.
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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 

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. 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).

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.

In 1958, Building 901W 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.
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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 26Mg(t,p)28Mg 

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).

2.4.10 National Synchrotron Light Source, Building 725

Building 725 houses 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 energy is 800 MeV. The radiation 

produced by the circulating electrons travels 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.

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 began operations in 1984 (BNL 1997l, pp. 1, 2).

Building 729 is the NSLS Source Development Laboratory. The building houses a LINAC coupled to 

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

2.4.11 Relativistic Heavy Ion Collider Complex, Building 1000 Series

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 beamline 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 there, 

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

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

described in Figure 2-5.

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

Building 1002

1981–present Experimental hall, assigned to the Broad Range Hadron Magnetic 

Spectrometers (BRAHMS) experiment

1982–1992 AGS storage, some activated components

1990–1992 AGS experiment 813

Building 1004A (Service Building)

1981–present Utilities, no process usage before 1992

1992–present RF system testing
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Figure 2-4. The Relativistic Heavy Ion Collider complex.

Building 1004 Open Area

1981–present Future experimental area

1994–present RF cavities testing for collider

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

Building 1004B (Support Building)

1994–present Power supplies, cryogenic valve boxes, other system electronics

Building 1005 Beam Tunnel

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

Building 1005S

1981-present Tech Shops on first floor, administrative offices on second, third, and fourth
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Figure 2-5. RHIC ring intersection points (BNL 2009b).

Building 1006

1991–present Solenoidal Tracker (STAR) Detector

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

1981–1990 Other short-term uses, temporary storage

Building 1006A (Service Building)

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

storage

Building 1007

1981–present Beam tunnel to be commissioned in 1999

1981–1995 Part of tunnel used as a calibration laboratory for Survey Group

Building 1008

1991–present Pioneering High Energy Nuclear Interaction Experiment (PHENIX)

1992–1995 Injection magnet fabrication

1988–1991 Uranium plate machining for the DØ [D-zero] Experiment at the Fermi National 

Accelerator Laboratory

Building 1008A

1981–present Experimental support building, utilities

1992–1995 Injection Magnet Group Shop/South Room

1990–1992 AGS/Physics Experiment Preparation/South Room

1986–1987 Assembly of wire chambers for AGS Experiment 755

1985–1986 Cleaning of depleted uranium with trichloroacetic acid for European 

Organization for Nuclear Research (CERN) Experiment NA-34
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Building 1010

1994–present PHOBOS Experiment2

Building 1012

1994–present Future Experimental Hall

2.4.12 Hot Laboratory, Building 801

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” area (the western portion) and a “Cold” area (the eastern 

side). 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 relatively high levels of radioactivity in 

gaseous, liquid, or solid form. The cells were maintained at negative pressure in relation 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 Laboratory (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 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, 

radiation-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 

 2 The name PHOBOS, which is a moon of the planet Mars, came about because the first proposed experiment was 

deemed too expensive and needed to be downscaled. It was called Modular Array for RHIC Spectra (MARS).
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used. An interviewee reported that at times the radioactive sample carriers, known as rabbits, 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. An explosion involving UF6 and BrF3 injured a few people and one was 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 

result of this incident. The floor is labeled as a radioactive material area (BNL 1997o, p. 7).

2.4.13 High Intensity Radiation Development Laboratory, Building 830

Building 830 is an office and laboratory complex presently occupied by Department of Applied 

Technology and RHIC personnel. All of the RHIC personnel are housed in a modular structure, which 

is all office space. The laboratory complex has always been used as an extensive experimental area, 

especially involving 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.

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 Chemistry Department, Building 555

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 
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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.

A large variety of research programs has 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).

Some radiological problems associated with the building are:

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

in stainless-steel containers that penetrate the laboratory floors and extend into the ground 

below. There is evidence that one 40-Ci source was leaking small amounts of 60Co into the 

water that surrounds it 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 nCi.)

• 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 Physics Department, Building 510

Building 510 was built in 1962 to house 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 also 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).

The northeast corner of the building houses the Health Physics Section (Rooms 1 to 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 
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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 Other Miscellaneous Radiological Research Facilities

2.4.16.1 Liquid Metal Fuel Reactor Support, Building 820

R&D 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, pp. 5, 55–

64). The simulated reactor was a utility test loop in a figure-eight design with molten bismuth-uranium 

metal in one loop and molten bismuth in the other loop. The bismuth-uranium metal included uranium 

at 1,000 ppm. The test loop was made of a 4-in.-diameter stainless-steel tube. The simulated reactor 

was heated by a 5-MW oil-fired boiler. Small amounts of fission products were generated in the 

experiments from the 235U in the simulated reactor. The support systems were used to study 

thermodynamic properties and removal of fission products from molten mixtures using nonradioactive 

salts of fission product elements. The facility also used X-ray machines to test welds in the 

experimental structures.

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).

2.4.16.3 Health Physics and Safety Instrumentation, Buildings 535 and 356

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 R&D 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 S&EP 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 thermoluminescent dosimeters (TLDs) in the 

building (first floor). The building also housed the equipment needed to operate the external 

monitoring program (X-ray machine, darkroom, densitometer, microscopes, etc.).

In 1978, the NSLS Department moved into the rear section of the basement and has occupied that 

space since. 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).

Another building operated by the instrument division was Building 356 (BNL 1997y, BNL 2006a). 

From 1960, an irradiation facility using 60Co sources was used to study the irradiation effects on 

semiconductor materials and devices, electronic systems, and optical materials. In 1986, ownership 
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of the facility was transferred from Physics to the Applied Sciences Department, and in 1991 a 20,000 

Ci 60Co source was installed.

2.4.16.4 Biology Department, Building 463

Building 463 was constructed in 1942 with several later additions to house the Biology Department 

(BNL 1997z). The building contains offices, wet laboratories, instrumentation rooms, and rooms for 

ancillary functions such as seminars, conferences, and a library. A Controlled Environmental 

Radiation Facility (CERF) is located in the sub-basement that houses strong gamma sources used for 

irradiation experiments. High activity 60Co and 137Cs sources have been in use in Building 463 since 

1951 (BNL 1997z, BNL 2006b).

2.4.16.5 Radiation Calibrations Facility, Building 348

Building 348 has been used by the S&EP Division since 1959 as a facility to repair and calibrate 

radiation detection instruments. Standard encapsulated radiation sources are used to provide beta, 

gamma, and neutron radiation fields for instrument calibration. In about 1997, a replacement facility 

was constructed next to the original building, also designated as Building 348 (BNL 1997aa, BNL 

2006c).

2.4.17 Waste Management Facility, Radwaste Reclamation Building, Building 865

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

operation activities. The new WMF replaced the original HWMF 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.

The WMF includes the Operations Building (Building 860), Resource Conservation and Recovery Act 

Waste Building (Building 855), Radwaste (Radioactive Waste) Reclamation Building (Building 865), 

and Mixed Waste Building (Building 870).

The Radwaste Reclamation Building (Building 865) 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, size-reduced, and packaged 

for temporary storage before shipment off site. To achieve the goal of waste reduction, Building 865 

houses a 500-lb/d lead smelter, a trash compactor, and a shredder (BNL 1997v, pp. 2, 3).

Additional radioactive waste treatment facilities include:

• Tritium Evaporator Facility (Building 802B). Wastewater processing began in 1995. This 

facility was constructed to reduce the total amount of tritiated water released to the Peconic 

River. Another room contains fans that exhaust Building 801 air to the stack.

• Waste Concentration Facility (WCF, Building 811. This facility 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 that is generated on the site, is 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 
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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.

• Waste Management Incinerator. The Laboratory incinerates certain low-level radioactive 

wastes at the HWMF incinerator, Building 444.

2.4.18 Radiological Waste Decontamination Facility, Building 650

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.

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- by 4- by 4-ft top tank used for acid, base, and 

detergent wash).

Associated with the Waste Decontamination Facility is an outside decontamination pad. From 1957 to 

1981, the 20- by 20-ft pad was used for washing 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 was characterized by OER in 1993 as containing 137Cs, 152Eu, 154Eu to a depth of 0.5 ft 

exceeding OER cleanup goals, and the soil is in the OER remedial plan. 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 2-1 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 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.

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 2-1 also provides a perspective of the magnitude of the nuclear operations at BNL. ORAUTPROC-0031, Site Profile and Technical Basis Document Development (ORAUT 2011a), requires the 
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magnitude of the operations be expressed in radioactivity (Ci) of the isotopes of concern. The same 

information is also required in Section 5.0, “Occupational Internal Dose.” Therefore, radioactivity 

values in Section 5.0 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 in the Section 5.0 table are the ratios of the activity of the 

individual released isotope and the total released activity of all the isotopes. In cases where numeric 

data are not available, the table provides descriptive text that gives a perspective on the magnitude of 

the radiological impact on personnel.

2.7 MAJOR INCIDENTS

Table 2-2 describes major site incidents that might have significant potential for internal or external 

exposure to personnel. The records include many incident investigations with estimates of dose to 

individuals. Most of these investigations involved localized contamination without release to the 

environment. 

2.8 RADIOLOGICAL ACCESS CONTROLS

Radiological access controls have been exercised locally at the site where activities involving 

radioactive materials have been performed. Radiological Control Areas have been established for 

each specific laboratory, facility, or building in which nuclear operations have been conducted at the 

BNL site. Entry to those areas has been controlled (i.e., hand-and-shoe counters, dosimeters, and 

training requirements).

Table 2-2. Major incidents.

Facility/

building Year Description

BGRR,

Building 701

1952–1957 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 from 1952 to 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).

BGRR, 

Building 703

12/20/1975 Contamination of Laboratory W-9 in Building 703 occurred during loading of a 

cerium sample. The contamination spread was identified and contaminated 

areas were decontaminated. Contaminated workers were provided bioassay 

monitoring. No release to the environment was identified (Roesler 1976).

Hot Laboratory,

Building 801

5/15/1957 There was a serious incident involving the Volatility Project where uranium 

reprocessing was the objective (Eisenbud 1957). 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. The UF6

combined with moisture to form an oxide. The BrF3 was 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 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 

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; the concrete is 
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Facility/

building Year Description

labeled as a radioactive area (BNL 1997o).

BLIP,

Building 931B

5/16/1978 A target at the BLIP facility vaporized causing radioactive material to be vented 

to the work environment and the general public. Bioassay monitoring was 

provided to two exposed individuals (Miltenberger 1978a). The estimated dose 

to the two workers was 0.275 rem and 1.350 rem to the thyroid, and the dose to 

the public was estimated to be 0.060 rem to the thyroid at the site boundary for 

full occupancy (Miltenberger 1978b). 

HFBR,

Building 750

3/18/1983 A rupture of the moderator chamber in the CNF occurred due to a rapid reaction 

of hydrogen and oxygen in the chamber. The incident resulted in loss of 

experimental time for researchers using the HFBR beam port H-9. No one was 

injured, and there was no impact to the public (BNL 1983).

3/31/1994 On March 31, 1994, smoke was detected emanating from the TRISTAN 

Experimental facility. The TRISTAN (Terrific Reactor Isotope To Analyze 

Nuclides) was an on-line isotope separator located on the experimental level at 

beam hole H-2 of the HFBR. Radioactive contamination in the air on the 

experimental level was detected. The reactor was shut down and the reactor 

building was evacuated. The portal monitors detected contamination on 7 

persons as they left the reactor building. The fire was most likely caused by 

electrical failure in the high current filament leads. The fire resulted in 

contamination of the reactor containment building with a small release to the 

outside (BNL 1994).

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 close the HFBR permanently (BNL 2001a).

HFBR,

Building 750 

(continued)

9/27/1999 Two gallons of D2O drained into a sump on the equipment level, partially filling it 

with D2O. The portable tritium monitor spiked at 1000 microcuries/m3. The offgas system evaporated the liquid. The three workers involved in the incident 

received estimated thyroid doses of from 71 to 265 rem from radioiodine 

isotopes, and the maximum estimate thyroid dose to an individual at the site 

perimeter was 60 mrem (BNL undated a)

3.0 OCCUPATIONAL MEDICAL DOSE

3.1 PURPOSE

Work funded by the U.S. Atomic Energy Commission (AEC) began at BNL in 1948. BNL occupational 

medical physicians prescribed chest X-rays as part of the preemployment and periodic physical 

examination of employees from the beginning of operations (Sunderman 1947a; Cowan 1948). 

These radiographs caused exposure of the lungs and other organs and tissues of the body. Exposure 

of the radiographed workers came mostly 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 about the historical BNL occupational medical X-ray program has been found. 

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 about the early radiographic examination practices, equipment, and 

exposure rates.
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3.3 EXAMINATION FREQUENCY

The occupational medical program at BNL began in 1947. Sunderman (1947a) stated that the 

recommended X-ray screening protocol included 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 included a PFG; and the 

recommended termination examination included a PFG, AP and LAT lumbar spine, and the same 

forearm radiographed in 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 

termination. These recommendations specifically referred to chest X-rays on 14-in. by 17-in. film 

(Wolf 1948).

The X-ray records in the claim files from BNL do not bear out the recommendations by Sunderman 

(1947a). 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 reviewed sample of cases [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, 

had a preemployment and an annual physical examination, including a PA chest X-ray on 14- by

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- by 17-in. PA chest 

X-rays in the claim file records if present. If these records are not present, the dose reconstructor 

should assign dose from a preemployment PFG for start dates between 1947 and 1955 and from 

annual 14- by 17-in. PA chests during this period. The default X-ray examination frequencies are 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, if PFG in the records, or no 

records exist

Periodic per the records

No termination examination

1947–1955 14- by 17-in. PA All workers Preemployment if no PFG in the records

Annual

No termination examination

1956–1978 14- by 17-in. PA All workers Preemployment

Annual

No termination examination

1979–present 14- by 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, 72). It is assumed this examination frequency 

pattern continued through 1978. Dose reconstructors should assign dose according the number of 

14- by 17-in. PA chests in the claim file records if present. If these records are not present, the dose 

reconstructor should assign dose from a preemployment 14- by 17-in. PA chest and from annual 14-

by 17-in. PA chests during this period (1956 to 1978). The default X-ray examination frequencies are 

in Table 3-1.
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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 previous 

clinical evaluation was made (Carter, J. 1978). This recommendation, at least in part, was 

incorporated into BNL practice, and the average time between radiographic examinations increased. 

Starting in 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. Sometime 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 have started in 1979 coinciding with the extended reexamination 

frequency. The 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 for 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).

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 (1947b) states, “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 wellknown 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 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 2011b), the dose values from that document should be used for dose 

reconstruction of PFG for BNL workers.

Sunderman (1947c) 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 about 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- by 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 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 equivalent inherent filtration, 0.5-mm Al added filtration, and peak 

kilovoltage of 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 
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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 

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 at 80 kVp (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 set the 

collimation manually. 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 INCIDENT AIR KERMA

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 

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, it is not likely to improve the accuracy of the dose reconstruction significantly.

1947 to 1955, PFG

PFG incident air kerma and organ doses are from ORAUT-OTIB-0006 (ORAUT 2011b).

1947 to 1959, 14- by 17-in. PA Chest

The incident air kerma in air for the 14- by 17-in. radiographs is 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 2011b). The calculation follows:

1. From NCRP Report 102 (NCRP 1989), Table B-3, the average air kerma rate for a singlephase X-ray machine with 2.5-mm Al equivalent 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 equivalent using the information in ORAUT-OTIB0006 (ORAUT 2011b, pg. 9), 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.
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 49 of 131

4. Correcting for a source-to-skin distance of 155 cm, the incident air kerma is 0.06 cGy. The 

measurement of 0.055-R reported by Handloser and Love (1951) supports this value.

1960 to 1978

The incident air kerma for this period is based on the measurements on X-ray equipment reported by 

Carter, N. (1978). Carter reports a measured ESE of 7 mR (on the “old” Picker machine) and 11 mR 

(on the “new” Picker machine) for the PA chest at 115 kV, 100 mA, 1/30 second, and a measured 

HVL of approximately 3.0 mm Al equivalent (Carter, N. 1978). Both the ESE and the unit of Roentgen 

are now considered obsolete. Therefore, if ESE is converted to incident air kerma (as described in 

equation 1 of ORAUT-OTIB-0006), then the resulting value (9.6 cGy) is slightly less than the 11 mR 

reported by Carter (and used in the previous revision of this TBD). Since the values are so close, 

Carter’s value was used as an incident air kerma for dose reconstruction.

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, in comparison with 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. It is reasonable to assume that the first three of five measurements, ranging from 15.6 to 

37.6 mR are for the PA chest projection, and the latter two, ranging from 50.4 to 73.0 mR are for the 

LAT chest projection. Both the ESE and the unit of Roentgen are now considered obsolete. 

Therefore, if ESE is converted to incident air kerma (as described in equation 1 of ORAUT-OTIB0006), then the resulting values are slightly less than the exposure values reported by Zukas (and 

used in the previous revision of this TBD). Since the values are so close, Zukas’s values were used

as incident air kerma values. For dose reconstruction, the incident air kerma for the PA chest for this 

period is assumed to be 0.030 cGy, and for the LAT chest it is assumed to be 0.060 cGy.

1987 to 1995

In 1987, measurements of the X-ray machine in Room 2 were made at 100 kVp and 200 mA and 

300 mA at various exposure times (Zukas 1987). The listed entrance exposures are lower than those 

determined in 1983, so it is possible there was a change in technique factors and/or screen-and-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. It is reasonable to assume that the first three of four measurements, 

ranging from 8 to 16.2 mR are for the PA chest projection, and the last one, at 35.6 mR is for the LAT 

chest projection. Both the ESE and the unit of Roentgen are now considered obsolete. Therefore, if 

ESE is converted to incident air kerma (as described in equation 1 of ORAUT-OTIB-0006), then the 

resulting values are slightly less than the exposure values reported by Zukas (and used in the 

previous revision of this TBD). Since the values are so close, Zukas’s values were used as incident 

air kerma values. For dose reconstruction, the incident air kerma for the PA chest for this period is 

assumed to be 0.015 cGy, and for the LAT chest it is assumed to be 0.030 cGy.

1996 to present

Between 1998 and 2004, the PA chest ESE from the Picker unit in Room 1 was measured at least 

four times by American Board of Radiology/American Board of Medical Physics-certified radiation 

physics consultants during the course of routine inspection (Astarita Associates 1998–2004). The 

technique factors were similar to those for nominal PA and LAT chest examination in current use, 

namely 110 kVp and 3.2 mAs on the large focal spot with a 72-in. source-to-image distance and 

23-cm phantom. The measured ESEs were approximately 7 mR. For 1996 to 2010, the incident air 

kerma for PA chest is rounded up to 0.010 cGy 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 projection. 

Therefore, the incident air kerma for the LAT chest was calculated to be 0.025 cGy based on the rule 

of thumb that the incident air kerma for LAT chest is conservatively estimated to be 2.5 times that of 

the PA chest (ORAUT 2011b, p. 22).
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 50 of 131

3.6 ORGAN DOSE CALCULATIONS

Organ doses for PFG, PA, and LAT chest X-rays were determined using the method described in 

ORAUT-OTIB-0006 (ORAUT 2011b) with dose conversion factors (DCFs) from ICRP Publication 34 

(1982). ICRP (1982) provides tables of average absorbed dose (in milligray) in selected organs for 

selected X-ray projections at 1-Gy entrance kerma (i.e., air kerma without backscatter) and selected 

beam qualities (i.e., various HVLs). The Publication 34 tables list the basic DCFs for converting air 

kerma to organ dose. Substitute DCFs for organs that are listed in the Interactive 

RadioEpidemiological Program (IREP) but without unique DCFs in ICRP Publication 34 (1982) were 

selected as described in ORAUT-OTIB-0006 or are footnoted in the organ dose tables. Incident air 

kerma was obtained from Table 3-2.

Table 3-2. X-ray equipment used at BNL and incident air kerma values.

Period X-ray machine

kVp

(kV)

Assumed HVL 

(mm Al) Image size 

Incident air kerma (cGy)

PA chest LAT chest

1947–1955 PFG -- 2.5 4 by 5 in.

or 70 mm

2.27a --

1947–1959 Westinghouseb 72b 1.5 14 by 17 in. 0.06 --

1960–1978 Pickerc 115c 3.0c 14 by 17 in. 0.011c --

1979–1986 -- 120d 3.0 14 by 17 in. 0.030 0.060

1987–1995 Pickerf 110e 3.0f 14 by 17 in. 0.015 0.030

1996–present Pickerf 110f 3.5 14 by 17 in. 0.010f 0.025

a. From ORAUT-OTIB-0006 (ORAUT 2011b). The reported incident air kerma is for stereo PFG views.

b. Handloser and Love (1951).

c. Carter, N. (1978).

d. Zukas (1983).

e. Zukas (1987).

f. Astarita Associates (1998–2004).

Table 3-3 lists the organ doses for all periods. Skin doses for all skin areas were determined 

according to the method described in ORAUT-OTIB-0006 (ORAUT 2011b) and listed in Table 3-5.
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 51 of 131

Table 3-3. Organ doses (rem) for PFG, PA, and LAT 14- by 17-in. chest radiography.

Organ Projection

1947–1955 

PFGa

1947–1959 

14- by 17-in. 1960–1978 1979–1986 1987–1995

1996 to 

present

Thyroid PA 3.94E-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 7.25E-02 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

Ovaries PA 2.50E-02 4.14E-03 1.98E-05 5.40E-05 2.70E-05 3.20E-05

LAT 5.40E-05 2.70E-05 4.00E-05

Urinary 

bladder/prostate

PA 2.50E-02 4.14E-03 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 4.14E-03 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 1.86E-04 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 9.50E-01 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 1.02E+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 1.02E+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 1.02E+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 1.02E+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 1.02E+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/ 

pancreas

PA 1.02E+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 1.02E+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 1.11E-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 3.84E-03 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 2.09E-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 1.95E-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 skinb PA 3.06E+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 2011b). The doses are for stereo PFG (i.e., two exposures).

b. Entrance skin dose is the incident air kerma in air multiplied by the backscatter factor of 1.30, 1.35, or 1.4 for HVLs of 

1.5, 2.5, or 3.0 mm Al, respectively, from NCRP Report 102 (NCRP 1989).

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 energy employee for screening and as 

a condition of employment. Only radiographs made for screening and as a condition of employment 

are eligible to be included in dose reconstructions. Such exposures are limited to PA and LAT chest 

radiographs and were required or recommended as part of a preemployment examination or a routine 

reexamination.
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 52 of 131

Table 3-4. Skin dose guidance for various chest projections and periods.

a

Area of skin PFG

Poorly 

collimated 

PA chest

Properly 

collimated 

PA chest

Properly 

collimated 

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

Right elbow, forearm, wrist

Right elbow, forearm, wrist

10% ENSD ENSD 10% ENSD 10% ENSD

Right side of head (including temple and ear) 10% ENSD 10% ENSD 10% ENSD 10% ENSD

Left side of head (including temple and 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

a. ENSD = entrance skin dose; EXSD = exit skin dose; RSD = remote skin dose.

Medical X-rays were recorded on a form called “X-ray Examination Request and Report” that was 

used for all X-rays including screening examinations, injury investigations, and for diagnosing 

conditions such as shortness of breath. While the forms changed over time, they typically included a 

section for the date of the request, the reason for the request, and the part of the body to be 

examined. The examinations that are to be considered in the dose reconstruction are those that 

indicate “Pre-Emp. Physical,” “Physical Recheck,” or “Termination” as the reason for the examination. 
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 53 of 131

Table 3-5. 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 6.67E-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 3.06E+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 6.67E-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 3.06E+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 3.06E-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 3.06E-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 3.06E+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 3.06E+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

3.06E-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

3.06E-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 temple and ear) 3.06E-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 temple and ear) 3.06E-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 9.E-04 1.E-05 5.E-06 1.E-05 1.E-05 7.E-06 6.E-06 5.E-06 6.E-06

Back left thigh 9.E-04 1.E-05 5.E-06 1.E-05 1.E-05 7.E-06 6.E-06 5.E-06 6.E-06

Front right thigh 9.E-04 1.E-05 5.E-06 1.E-05 1.E-05 7.E-06 6.E-06 5.E-06 6.E-06

Back right thigh 9.E-04 1.E-05 5.E-06 1.E-05 1.E-05 7.E-06 6.E-06 5.E-06 6.E-06

Left knee and below 3.E-04 4.E-06 2.E-06 5.E-06 5.E-06 2.E-06 2.E-06 2.E-06 2.E-06

Right knee and below 3.E-04 4.E-06 2.E-06 5.E-06 5.E-06 2.E-06 2.E-06 2.E-06 2.E-06

Left side of face 7.25E-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 7.25E-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 3.06E-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 3.06E-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 3.06E-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 7.25E-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 3.06E-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 6.67E-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 6.67E-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 6.67E-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 6.7E-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 3.06E+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 3.06E+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 3.06E+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

Back torso: buttocks (Iliac crest and below) 3.06E-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 3.06E+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 3.06E+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
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 54 of 131

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 torso: lowest rib to iliac crest 3.06E+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) 3.06E-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 3.06E+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 3.06E+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 3.06E+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) 3.06E-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 shown to nearest 0.1 mrem
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 55 of 131

The part of the body examined will be given as “Chest” or “Routine Chest.” The presence of a request 

for X-ray examination does not mean the employee was actually given the examination. After about 

1979, there may be handwritten entries of “Not Indicated” or “Declined” that indicate the X-ray 

examination was not given and no dose should be assigned. The form (usually the bottom part) 

indicates the date the examination was given and includes an interpretation (report) by a radiologist. 

If the form does not include a date given and an interpretation, then it is unlikely the examination was 

actually given; therefore, dose should not be assigned. In the early years (1947 to 1955), the record 

will likely indicate if the X-ray was a PFG (identified as “photofluoro”) or a “14 x 17” PA film.

If confusion about the radiographic exposure record exists, the dose reconstructor should 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. The dose 

reconstructor should assume that any radiograph that was not a PA or LAT chest was performed as a 

diagnostic test rather than screening and is therefore 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). The dose 

from these radiographs is not included in dose reconstruction.

The photon energy associated with all occupational medical X-ray dose is in the 30- to 250-keV 

energy range.

3.8 UNCERTAINTY

ORAUT-OTIB-0006 (ORAUT 2011b) lists the major sources of uncertainty in X-ray output intensity 

and subsequent effect on dose to the worker. The five sources of uncertainty are:

1. X-ray beam measurement error (±2%),

2. Variation in peak kilovoltage (±9%),

3. Variation in X-ray beam current (±5%),

4. Variation in exposure time (±25%), and

5. Variation in source-to-skin distance as a result of worker size (±10%).

The 10% uncertainty in output intensity as a result of worker size was based on an inverse square 

correction of output intensity changes from differences of standard chest thickness of ±7.5 cm.

These uncertainties are assumed to be random; therefore, the combined statistical uncertainty was 

calculated as the square root of the sum of the squares of all the uncertainties, which is ±28.9%. 

Rounding this up to ±30% provides an adequate and suitably conservative indication of uncertainty. 

Therefore, for a derived dose equivalent to an individual organ, a total combined standard uncertainty 

of ±30% can be assumed. Dose reconstructors should, therefore, input the organ dose equivalent as 

the mean of a normal distribution with a standard deviation of 30%.

4.0 OCCUPATIONAL ENVIRONMENTAL DOSE

4.1 INTRODUCTION

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 
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 56 of 131

assemblies and 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 around the BNL site at the northwest, southwest, southeast, and northeast.

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 Onsite Monitoring

Table 4-1 provides average and maximum external gamma radiation results that have been adjusted 

to reflect a 2,000-hr/yr occupancy. The table also provides annual dose rates at normally unoccupied 

locations near sources of radiation on the site. The basis for the values in this table are discussed in 

this section. The following references were used to develop the values in the table: BNL 1963a; Hull 

1963, 1964, 1966; Hull and Gilmartin 1967, 1969; Hull 1970a,b,c; BNL 1972; Hull 1973; Gilmartin 

1974a through 1974l, Hull and Ash 1974 through 1976; Naidu 1977 through 1980; Gilmartin 1980a 

through 1980l; BNL 1980; Naidu and Olmer 1981, 1982; Gilmartin 1983a through 1983j, Day and 

Naidu 1983, 1984, Day, Miltenberger, and Naidu, 1985, 1986, Gilmartin 1986a through 1986e, 

Miltenberger, Royce, and Naidu 1987 through 1990, 1992; Naidu, Royce, and Miltenberger 1992, 

1993; Naidu and Royce 1994, 1995; Naidu et al. 1996; Meinhold and Hull 1998; Schroeder et al. 

1998, Lee et al. 1999; BNL 1999a, 2000; Meinhold and Meinhold 2001; BNL 2001a, 2002 to 2008, 

2009c, 2011.

Table 4-1. External gamma radiation dose (mrem/yr).

Year

Onsite locations Normally unoccupied locations

Average Maximum

Monitoring 

location P-9a HWMFbS-6

Building

356c Building913Bd

1947e 0

1948 0

1950f 0.18 0.27

1951 6.4 9.1

1952 8.0 11

1953 8.4 12

1954 8.9 13

1955 8.9 13

1956 8.4 12

1957 7.3 11

1958 21 30
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Year

Onsite locations Normally unoccupied locations

Average Maximum

Monitoring 

location P-9a

HWMFb

S-6

Building

356c

Building

913Bd

1959 27 39 10

1960 31 51 11

1961 28 38 8.8

1962 28 37 31

1963 28 42 30

1964 34 41 31

1965 44 62 29

1966 53 103 30

1967 55 110 23

1968 55 110 21

1969 55 110 19

1970 15 30 19

1971 15 30 3.7

1972 15 30 3.6

1973 15 30 3.5 202

1974 15 30 20 338

1975 15 30 1.1

1976 15 30 1.0

1977 15 30 17 190

1978 15 30 14 192

1979 15 30 11 184

1980 5 10 190

1981 5 10

1982 5 10

1983 5 10 213

1984 5 10

1985 5 10

1986 5 10 209

1987 5 10

1988 5 10

1989 5 10

1990 5 10

1991 5 10

1992 5 10 125

1993 5 10 105

1994 5 10 59

1995 5 10 73

1996 5 10 31

1997 5 10 37

1998 5 10 44

1999 5 10 33 5.0

2000 5 10 42 4.1

2001 5 10 25 5.9

2002 5 10 23 7.3

2003 5 10 24 16 2

2004 5 10 24 11 16

2005 5 10 5.3 6.4 16

2006 5 10 4.3 8.7 16

2007 5 10 4.6 12 11

2008 5 10 4.1 11 3
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Year

Onsite locations Normally unoccupied locations

Average Maximum

Monitoring 

location P-9a

HWMFb

S-6

Building

356c

Building

913Bd

2009 5 10 4.6 8.2 4

2010 5 10 4.6 53 4

a. Monitoring location P-9 was originally known as E-9. After 1979, the P-9 values are 

included in the average values.

b. The HWMF is represented by dosimeter location S-6. Monitoring data for this site are 

sporadic.

c. Building 356 contains a Co-60 source. The only monitoring data available starts in 1999.

d. Building 913B dose values represent areas near the AGS tunnel access. The only 

monitoring data available starts in 2003.

e. Data before 1950 are assumed based on site activities in those years.

f. Data from 1950 to 1961 are from Meinhold and Meinhold (2001).

The site performed environmental monitoring of ambient external radiation levels starting in 1949. 

Initially there were 16 monitoring stations to measure the ambient levels before and after startup of 

the BGRR. The locations included onsite locations, perimeter locations, and offsite locations. Four 

onsite monitoring locations (S-10, S-11, S-12, and S-13) were use to measure external dose from 41Ar

releases from the BGRR and later the HFBR. Four perimeter locations were used to monitor for 

external dose to demonstrate the maximum potential dose for offsite locations. Both of the reactors 

released activity through the same stack (Meinhold and Meinhold 2001, p. 33). During the first 

decade of operation, the BGRR was fueled by natural uranium and released about 7,000 Ci/d of 41Ar. 

In 1958, the fuel was changed to enriched uranium and the emission rate increased to about 

20,000 Ci/d. The measured doses above background levels were found to be directly related to the 

41Ar emission rate. BGRR operations ended in 1969. The HFBR released much lower levels of 41Ar

with a corresponding smaller contribution to onsite and perimeter external dose.

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, in addition to four perimeter locations.

From 1961 through September of 1979, a large 137Cs source was used to expose a forested area in 

the Northeast area of the site. The site annual reports during this period list the external dose 

contribution from this source to the northeast perimeter location and to one of the onsite locations 

(S-13).

In 1980, the environmental monitoring program was expanded by adding 40 additional TLD 

monitoring stations in rings at 1, 3, and 10 km from the HFBR stack (Gilmartin 1980a). However, 

results from the added onsite monitoring stations are not available until 1985. The 1981 annual 

environmental report (Naidu and Olmer 1982, p. 22) indicated that the Northeast perimeter location 

(P-9) lies on a bed of coal cinders that contain radium and thorium at larger concentrations than the 

foundation material used at other perimeter stations, which results in a background that is slightly 

different from the other perimeters stations. However, the annual external dose values for the 

perimeter locations in Table 4-1 are not adjusted for this effect.

The HWMF was used to store highly radioactive materials in shielded vaults, trenches, and vertical 

tubes in the ground. The facility has been in operation since the early 1950s. Monitoring station S-6 

is located directly opposite the HWMF area, where low and intermediate level radioactive materials 

are stored before offsite disposal. The annual external doses for this monitoring station are available 

for some but not all years. This area includes the igloo storage structures. In 1992 an effort was 

made to reduce the skyshine from these sources to nearby facilities (Van Der Karr 1993). The HWMF 

includes several small buildings, only one of which is normally used by workers. This is Building 445,

which was constructed in 1956 and housed a triple distillation still for mercury until 1980 (BNL 1997x). 
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 59 of 131

From 1981 to 1991 the building housed the waste management incinerator. Other buildings in the 

area were used only for storage and were not occupied on a regular basis. As part of the 1992 effort 

to reduce external dose in the HWMF area, surveys were taken before and after the relocation of 

vaults and the addition of shielding to radiation sources. The highest annual dose rate (for 2,000-hr/yr

occupancy) in the vicinity of Building 445 was found to be 125 mrem/yr before the source relocation 

and 105 mrem/yr afterward. These values are based on monitoring for only a 2-month period, before 

and after the relocation. Most of the radioactive materials were removed from the HWMF by the end 

of 2004, and the doses decreased to less than 10 mrem/yr.

Another potential source of onsite exposure is Building 356 (BNL 1997y, 2009a). The building housed 

a 60Co irradiation facility under the Physics Department from 1960 to 1966, which was 

decommissioned in 1965 and 1966. During 1964 to 1966 an addition to the building was designed 

and built to house the present irradiation facility. The facility was upgraded in 1991 with a new 

20,000-curie 60Co source. The areas outside the building were added to the environmental monitoring 

program, and doses were reported starting in 1999. The annual doses for this location were generally 

well under 15 mrem/yr, except for 2010 when the dosimeter at the North corner of the building 

indicated 53 mrem/yr (for 2,000-hr/yr occupancy). Although it is conceivable that individuals using the 

parking lot at Building 356 could receive a dose from this source, the annual dose would be small 

because of the short occupancy period for the location (BNL 2009a).

In about 1990, dosimeter locations were added to monitor potential external exposure points near 

operating facilities with the potential to deliver radiation dose in occupied areas and unoccupied 

areas. By 1999 these locations included several near the AGS, BLIP, and by 2001 locations near the 

new RHIC. The recorded doses for all of these locations are included in the summary information in 

Table 4-1 and indicate that average annual dose (based on 2,000-hr/yr occupancy) is well under 

5 mrem/yr from 1984 through 2010 and the maximum dose for occupied areas is under 10 mrem/yr.

Starting in 2003, two dosimeter locations near the AGS tunnel access were added to monitor skyshine 

from the AGS (BNL 2006d). The maximum of the two dosimeters was always less than 20 mrem/yr. 

These locations would normally be unoccupied.

In 2010, neutron monitors were placed around the RHIC, AGS, and BLIP facilities. The highest 

monitored neutron dose was 1 mrem/quarter near these facilities, which indicates an annual dose of 

less than 1 mrem/yr based on 2,000-hr/yr occupancy. Therefore, the assignment of ambient neutron 

dose is not necessary.

The available information covers most of the site operating period except for 1967 through 1983, 

although some information is available for some of these years. For 1967 through 1969 (when the 

BGRR was still running), the values for 1966 are used to estimate the average and maximum dose. 

This likely represents an overestimate of dose values because the 1966 values are the highest up to 

that point. For 1970 through 1983, the average dose is set to 15 mrem/yr based on the available 

perimeter and onsite station S-13 values. The maximum is set to twice the average value, based on 

the 1966 ratio of the maximum dose to average dose (the year of the highest ratio for all previous 

years). For 1984 through 2010, the average values were all under 5 mrem/yr and the maximum 

values were all under 10 mrem/yr. Therefore, these values have been used as the average and 

maximum values for all of these years.

4.4.2 Applicability for Dose Reconstruction

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. The average dose values from Table 4-1 (based on exposure for 2,000 hr/yr) 

should be assigned to unmonitored workers as a normal distribution with a standard deviation of 
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 60 of 131

100%. This assumes the maximum value (2 times the average value) represents 1 standard 

deviation. This is reasonable considering the relatively small number of onsite monitoring locations 

for most years. The values should be corrected for annual exposure time as necessary for each year 

of exposure.

The dose values in Table 4-1 for unoccupied locations should not normally be applied for unmonitored 

workers because workers in these locations would normally be assigned radiation dosimeters. If the 

available information indicates a worker was assigned to one of these areas, and no dosimetry 

records exist, then the assignment of unmonitored doses described in Section 6.5.2 should be 

considered.

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 Stack Monitoring

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 1.2 m3

/hr (ICRP 1994) based on a 2,000-hr/yr 

exposure.

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 1966–2010 BMRR 1996–1997

Hot Lab & HFBR 1986 Incinerator 1981–1996

Hot Lab 1987–2004 Evaporator 1995–present

BLIP/LINAC 1973–2010

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 Estimates of Potential Inhalation Intake

The intake of each radionuclide that substantially contributes to personnel dose is tabulated by year 

for 1950 through 2004. After 2004, the only radionuclide of significance identified in the airborne 

effluent was tritium. In Table 4-2, the maximum annual intakes are provided for 1950 through 2004. 

Intake values for tritium for 2005 to 2010 are given in Table 4-3. For other radionuclides after 2004, 

use the 2004 values from Table 4-2. These values are then assigned as a lognormal distribution with 

a geometric standard deviation (GSD) of 3.

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 
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Table 4-2. Maximum annual environmental occupational radionuclide inhalation (Bq/yr).

a,b,c

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-99d

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-99d

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-99me Ru-103

Rh-106

Ru-106 Cs-137 Ba-140 La-140 Ce-141 Ce-144 Ra-226 Th-228

1950 N/A 6.51E+01 5.27E+01 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1951 N/A 6.51E+01 5.27E+01 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1952 N/A 6.51E+01 5.27E+01 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1953 N/A 6.51E+01 5.27E+01 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1954 N/A 6.51E+01 5.27E+01 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1955 N/A 6.51E+01 5.27E+01 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1956 N/A 6.51E+01 5.27E+01 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1957 N/A 6.51E+01 5.27E+01 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1958 N/A 6.51E+01 5.27E+01 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1959 9.82E+03 6.51E+01 5.27E+01 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1960 9.82E+03 6.51E+01 5.27E+01 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1961 9.82E+03 6.51E+01 5.27E+01 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1962 9.82E+03 6.51E+01 5.27E+01 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1963 9.82E+03 6.51E+01 5.27E+01 2.02E+01 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1964 9.82E+03 8.79E-01 1.23E+01 7.03E+00 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1965 9.82E+03 8.79E-01 2.20E+00 2.37E+00 1.23E+00 1.23E+00 8.79E-01 6.33E+01 2.55E-01 1.76E-01

1966 9.82E+03 6.15E-01 5.27E-01 7.03E-01 1.23E+00 1.23E+00 4.40E-01 6.33E+01 2.55E-01 1.76E-01

1967 9.82E+03 1.05E+00 5.27E-01 7.03E-01 1.23E+00 1.23E+00 4.40E-01 6.33E+01 2.55E-01 1.76E-01

1968 9.82E+03 1.05E+00 5.27E-01 6.15E-01 5.27E-01 5.27E-01 4.40E-01 6.33E+01 2.55E-01 1.76E-01

1969 9.82E+03 1.05E+00 2.11E+00 5.27E-01 1.76E-01 1.76E-01 4.40E-01 1.58E+01 2.55E-01 1.76E-01
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Year

Radionuclide

Tc-99me Ru-103

Rh-106

Ru-106 Cs-137 Ba-140 La-140 Ce-141 Ce-144 Ra-226 Th-228

1970 9.82E+03 1.05E+00 2.11E+00 1.23E+00 1.76E-01 1.76E-01 4.40E-01 2.46E+00 2.55E-01 1.76E-01

1971 9.82E+03 1.05E+00 2.11E+00 1.05E+00 8.79E-02 8.79E-02 4.40E-01 6.15E-01 2.55E-01 1.76E-01

1972 9.82E+03 1.05E+00 2.11E+00 4.40E-01 2.64E-01 2.64E-01 4.40E-01 2.37E+00 2.55E-01 1.76E-01

1973 9.82E+03 1.05E+00 1.23E+00 8.79E-02 2.64E-01 2.64E-01 4.40E-01 2.37E+00 2.55E-01 1.76E-01

1974 9.82E+03 1.05E+00 1.23E+00 8.79E-02 2.64E-01 2.64E-01 4.40E-01 4.22E+00 2.55E-01 1.76E-01

1975 9.82E+03 1.05E+00 6.15E-01 8.79E-02 3.52E-01 2.64E-01 4.40E-01 6.68E+00 2.55E-01 1.76E-01

1976 9.82E+03 4.88E+03 6.15E-01 8.79E-02 3.52E-01 3.52E-01 4.40E-01 8.00E+00 2.55E-01 1.76E-01

1977 9.82E+03 4.88E+03 7.91E-01 8.79E-01 3.52E-01 3.52E-01 3.52E-02 1.32E+00 2.55E-01 1.76E-01

1978 9.82E+03 4.88E+03 7.91E-01 8.79E-01 3.52E-01 3.52E-01 3.52E-02 3.52E-01 2.55E-01 1.76E-01

1979 9.82E+03 4.88E+03 7.91E-01 8.79E-01 2.20E-01 2.20E-01 3.52E-02 2.02E+00 2.55E-01 1.76E-01

1980 9.82E+03 4.88E+03 7.91E-01 8.79E-01 2.20E-01 2.20E-01 3.52E-02 7.91E-01 2.55E-01 1.76E-01

1981 9.82E+03 4.88E+03 7.91E-01 3.96E-01 2.20E-01 2.20E-01 3.52E-02 7.91E-01 2.55E-01 1.76E-01

1982 9.82E+03 4.88E+03 9.67E-02 2.64E+00 2.20E-01 2.20E-01 3.52E-02 5.36E-03 2.55E-01 1.76E-01

1983 9.82E+03 4.88E+03 9.67E-02 2.99E-01 2.20E-01 2.20E-01 3.52E-02 5.36E-03 2.55E-01 1.76E-01

1984 9.82E+03 4.88E+03 9.67E-02 1.77E-01 3.70E-03 2.20E-01 3.52E-02 5.36E-03 2.55E-01 1.76E-01

1985 9.82E+03 5.86E+02 9.67E-02 2.81E-01 3.70E-03 2.20E-01 3.52E-02 5.36E-03 2.55E-01 1.76E-01

1986 9.82E+03 5.86E+02 7.03E-02 2.81E-01 3.70E-03 2.20E-01 3.52E-02 5.36E-03 2.55E-01 1.76E-01

1987 4.88E+02 5.86E+02 7.03E-02 2.17E-01 3.70E-03 2.20E-01 3.52E-02 5.36E-03 2.55E-01 5.27E-02

1988 4.88E+02 1.12E+00 7.03E-02 2.00E-01 3.70E-03 1.86E+02 3.52E-02 5.36E-03 8.44E-01 8.79E-02

1989 7.33E+03 1.12E+00 1.28E+02 4.48E-01 3.70E-03 1.86E+02 3.52E-02 5.36E-03 6.15E-01 8.79E-02

1990 7.33E+03 1.52E+01 1.28E+02 6.15E-01 3.70E-03 1.86E+02 3.52E-02 1.86E+00 1.03E+00 3.52E-02

1991 7.33E+03 1.52E+01 1.28E+02 7.77E-01 3.70E-03 1.86E+02 3.52E-02 1.86E+00 1.86E-01 3.52E-02

1992 7.33E+03 1.52E+01 1.28E+02 2.50E-01 7.76E+01 1.86E+02 9.41E-01 1.86E+00 1.86E-01 1.55E-01

1993 7.33E+03 1.52E+01 1.28E+02 4.55E-01 7.76E+01 1.86E+02 9.41E-01 1.86E+00 1.86E-01 1.55E-01

1994 3.03E+02 1.52E+01 1.28E+02 1.38E+00 7.76E+01 1.86E+02 9.41E-01 1.86E+00 5.07E+00 1.55E-01

1995 3.03E+02 1.52E+01 1.28E+02 2.76E+01 7.76E+01 1.86E+02 9.41E-01 1.86E+00 5.07E+00 1.55E-01

1996 3.03E+02 1.52E+01 1.28E+02 5.86E-01 7.76E+01 1.86E+02 9.41E-01 1.86E+00 5.07E+00 1.55E-01

1997 3.03E+02 1.52E+01 1.28E+02 1.47E-01 8.07E+02 4.27E+03 9.41E-01 7.34E+00 5.07E+00 1.55E-01

1998 3.03E+02 1.52E+01 1.28E+02 1.31E-02 8.07E+02 4.27E+03 9.41E-01 7.34E+00 5.07E+00 1.55E-01

1999 3.03E+02 1.52E+01 1.28E+02 2.59E-01 8.07E+02 4.27E+03 9.41E-01 7.34E+00 5.07E+00 1.55E-01

2000 3.03E+02 1.52E+01 1.28E+02 4.25E-02 8.07E+02 4.27E+03 9.41E-01 7.34E+00 5.07E+00 1.55E-01

2001 3.03E+02 1.52E+01 1.28E+02 4.25E-02 8.07E+02 4.27E+03 9.41E-01 7.34E+00 5.07E+00 1.55E-01

2002 3.03E+02 1.52E+01 1.28E+02 5.44E-04 8.07E+02 4.27E+03 9.41E-01 7.34E+00 5.07E+00 1.55E-01

2003 3.03E+02 1.52E+01 1.28E+02 5.44E-04 8.07E+02 4.27E+03 9.41E-01 7.34E+00 5.07E+00 1.55E-01

2004 3.03E+02 1.52E+01 1.28E+02 1.81E-09 8.07E+02 4.27E+03 9.41E-01 7.34E+00 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. Other radionuclides were reported, but were not included because the maximum dose was less than 0.001 rem to the highest metabolic 

organ.

d. Tc-99 reported only for incinerator, which operated from 1981 to 1996.

e. Tc-99m is reported for the BMRR, which began operating in 1959; Tc-99m is also reported for incinerator.
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Table 4-3. Maximum annual environmental 

occupational radionuclide inhalation, 2005 

to 2010 (Bq/yr).

Year H-3

2005 1.31E+05

2005 2.95E+04

2007 9.74E+03

2008 1.47E+05

2009 2.80E+05

2010 3.98E+04

sampling location during the year, the assumed breathing rate, and an exposure time of 2,000-hr/yr. 

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.

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].
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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.

5.0 OCCUPATIONAL INTERNAL DOSE

5.1 INTRODUCTION

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 OCAS-IG-002, 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 program and evaluates the uncertainty for BNL exposure and dose records.

NIOSH has determined, and the Secretary of the U.S. Department 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, 1993 (other than tritium after December 31, 1964) (see 

Section 1.2).

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, 1993, 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.
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5.4 BIOASSAY RECORDS IN THE INDIVIDUAL FILES

The work performed at BNL included a large variety of operations with varying potential for internal 

exposure. The greatest potential for worker internal exposures at BNL were with the operating 

reactors (BGRR, HFBR, and BMRR) where exposure to airborne tritium, activation products, fission 

products, and fuel components was possible. Workers at the reactors were monitored for internal 

exposure with bioassay monitoring on a regular basis. Workers in some of the other facilities were 

also monitored if the there was a significant potential for internal exposure. The potential for worker 

internal exposure in accelerators and colliders was considered to be small because much of the 

activation isotopes were short lived or fixed in structural materials (Lessard, Schaefer, and Karol 

2001). However, the need for monitoring was not applied uniformly across the site and the bioassay 

monitoring records are not always available in the employee files. Data in worker files may contain 

results for bioassay monitoring as described in this section.

5.4.1 Bioassay Record (Form 1720)

The BNL bioassay record, Form 1720, contains one line (record) per sample. The record includes the 

following fields:

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,

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 mixed 

fission products (MFPs) in dpm/d beta activity, 90Sr in dpm/d 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 Tritium Exposure Evaluation Form

This form was used to calculate the WB 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:

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, and

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 
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 67 of 131

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 if the BNL Bioassay Record (Form 1720) discussed above is 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 Summary of Whole-Body Counting Results

Summaries of the WB 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].

5.4.4 Individual Whole-Body Counting Results

By the 1980s, WB 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:

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),

4. CONCENTRATION (UNITS) (activity), and

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) (activity), and

5. %ERROR (2 SD) (error).

Initially the ABACOS printouts displayed “<” and the 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 (BNL 

2001b). This was corrected when the software was upgraded and subsequently both a decision level 

and an MDA were reported [13].

5.4.5 Voluntary Intakes of Radionuclides

The Medical Department conducted research involving internal exposures to radionuclides by human 

volunteers (DOE 1995). The first experiment started in 1950 and the last started in 1976. Some 

experiments continued beyond this date as indicated by employee records that show internal 

exposures as late as 1985. While many of the experiments involved patients who were not 

employees at the laboratory, some involved volunteer employees. The experiments included a variety 

of radionuclides administered through injection, inhalation, and ingestion.
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The medical records for the energy employee should be reviewed for indications of participation in the 

voluntary exposure experiments. If information in the telephone interview record, DOL records, or 

DOE records indicates the employee participated in these experiments but the actual intakes are not 

included in the records, the dose reconstructor should request the voluntary study medical records 

from the MRC at BNL. The voluntary intakes are provided in a summary form that indicates the date, 

radionuclide administered, the mode of administration, and the quantity involved. 

5.5 IN VITRO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PROTOCOLS

5.5.1 History of In Vitro Urine Analysis

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 Measurement Laboratory (EML). In 2002, it 

became part of the U.S. Department of Homeland Security.

Initially, support was also obtained from Oak Ridge National Laboratory to perform urinalysis for 

plutonium, polonium, MFPs, barium-lanthanum, yttrium, and strontium. By 1951, BNL was doing its 

own R&D 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 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 1963b).

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 approximately April 2004. Due to some problems with meeting the 

contractual MDAs, Eberline was replaced by the Severn Trent Services Richland Laboratory (STRL) 

in December 2001. STRL also did the 3

H analyses when they were discontinued at the site. About 

August of 2007, STRL became TestAmerica Richland Laboratory (TARL). The current BNL bioassay 

program is accredited by DOELAP.

During the operation of the BGRR, some workers apparently were routinely sampled for MFPs 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. 
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 69 of 131

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 b, p. 14).

During most of the site’s existence, it appears that department or division safety representatives or 

managers designated personnel for participation in the internal dosimetry program [whole-body 

counts (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.

5.5.2 Fecal Sample Analysis

Although fecal sampling might have been used in response to specific events, no evidence of a 

comprehensive fecal sampling program was located.

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 found. 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.

24-hr urine samples for fission, 

activation, uranium, and 

transuranic elements

No comprehensive documentation was found. 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

WB 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.
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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-cm-thick 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 S&EP Division. In 1980, the MDL 

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 

microcuries by:

4.65

900 37

Ctb MDL

K = × ×

(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).

The shadow shield was replaced by a standup counter, but the date of this transition was not found. 

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:

MDA 3.29 3 Sb

KT

× + = (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; Sun, Reciniello, and Sengupta 2004).
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 71 of 131

5.6.2 Chest Counting

Chest counts were not done routinely until about 1993 for reactor workers (based on review of 

bioassay records). 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 did iodinations was on a thyroid counting program in 2006. 

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 Building 750, and termination of service for 

Reactor Division and S&EP personnel with offices in the 

HFBR building. 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 the site 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 TARL

contract laboratory reports results with both a 1-sigma counting error and a 1-sigma total error [14].

5.8 DETECTION LIMITS

Table 5-3 lists the MDAs for in vitro bioassay analyses, and Table 5-4 lists the MDAs for in vivo

bioassay analyses. The MDAs are listed 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. The values in Tables 5-3 and 5-4 are 

primarily from worker bioassay record files. When bioassay results are found for a worker, the 

detection limits are usually included in the report. For the period 1992 – 1999, the records may 

indicate the counting configuration was for a lung count. However, the actual configuration was 

consistent with a whole body count, and the results should be treated as a whole body count.

Table 5-3. Detection limits for in vitro bioassay.

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

Am-241 (STRL–TARL) Fecalc 2001 0.16 pCi/24 hr

Br-77 Urinalysis 1975 5 μCi/Ld

Cd-109 Urinalysis 1978 140 pCi/L

Co-58 Urinalysis 1994 2.0× 10-8 μCi/mLe

Co-60 Urinalysis 1978 14 pCi/L

Co-60 (STRL–TARL)

f Urinalysis 2002 10 pCi/L

Cs-132 Urinalysis 1978 26 pCi/L
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 72 of 131

Radionuclide

Method/

description Period MDAa

Reporting 

levelb

Cs-134 Urinalysis 1978 15 pCi/L

Cs-137 Urinalysis 1964–1967 0.001 μCi/L

Cs-137 Urinalysis 1978 14 pCi/L

Cs-137 Urinalysis 1994 25 pCi/L

Cs-137 (STRL–TARL) Urinalysis 2002 10 pCi/L

H-3 (in house lab)g Urinalysis 1959–1993 2.0 × 10-2 μCi/L

H-3 (in house lab)h Urinalysis 1994–2000 5.0 × 10-3 μCi/L

H-3 (required by contract) Urinalysis 2001–11/2004 1.0 × 10-2 μCi/L

H-3 (STRL–TARL) Urinalysis 12/2004–9/2008 5.0 × 10-3 μCi/L

I-126 Urinalysis 1978 42 pCi/L

I-131 Urinalysis 1978 13 pCi/L

MFPsg Urinalysis 1952–1956 5.0 dpm/24 hr

MFPs Urinalysis 1957–1976 2 dpm/24 hri

Na-22 Urinalysis 1978 19 pCi/L

P-32 Urinalysis 1957 0.02 μCi/24 hr

P-32 Urinalysis 1985 2000 pCi/L

Pd-109j Urinalysis 1982 1.6 pCi/mL

Po-210 Urinalysis 1961–1962 12 dpm/d

Pu-238 (Eberline) Urinalysis 1999–2000 6.0 × 10-2 pCi/L No threshold

Pu-238 (STRL–TARL)

f Urinalysis 2001–2008 6.0 × 10-2 pCi/L No threshold

Pu-238 (STRL–TARL) Fecalc 2001 0.10 pCi/24 hr No threshold

Pu-239 (ORNL) Urinalysis 1950–1952 0.1 dpm/24 hr

Pu-239 Urinalysis 1963–1971 0.05 dpm/24 hr

Pu-239/240 (Eberline) Urinalysis 1999–2000 6.0 × 10-2 pCi/L No threshold

Pu-239 (STRL–TARL) Urinalysis 2001–2006 6.0 × 10-2 pCi/L No threshold

Pu-239 (STRL–TARL) Fecalc 2001 0.081 pCi/24 hr No threshold

Pu-242 (STRL–TARL) Urinalysis 2004–2006 1.0 pCi/L No threshold

Pu-242 (STRL–TARL) Urinalysis 2007–2008 6.0 × 10-2 pCi/L No threshold

Ra-226 Urinalysis 1957 1.2 dpm/d

Ra-226 (STRL–TARL) Urinalysis 2003–2007 1.0 × 10-1 pCi/L No threshold

S-35 Urinalysis 1985 1800 pCi/L

Sr-90d Urinalysis 1950–1998 5.0 dpm/24 hr

Sr-90 (Eberline) Urinalysis 1999–2000 1.0 pCi/L No threshold

Sr-90 (STRL–TARL) Urinalysis 2001–2008 1.0 pCi/L No threshold

Sr-90 (STRL–TARL) Fecalc 2001 6.2 × 10-2 pCi/24 hr No threshold

Th-228 (STRL–TARL) Urinalysis 2001–2009 0.1 pCi/L No threshold

Th-230 (STRL–TARL) Urinalysis 2001–2009 0.1 pCi/L No threshold

Th-232 (STRL–TARL) Urinalysis 2001–2009 0.1 pCi/L No threshold

Uranium (fluorometric) Urinalysis Up to 1955 10 μg/L

Uranium (fluorometric) Urinalysis 1956–1958 1 μg/L

Uranium (fluorometric) Urinalysis 1959–1969 5 μg/L

U-238 (fluorometric–ORNL) Urinalysis 1988 1 μg/L

Uranium (radiometric) Urinalysis 1961–1969 0.2 dpm/d

U-233 (STRL–TARL) Urinalysis 2002 3.0 × 10-2 pCi/L No threshold

U-234 (Eberline) Urinalysis 1999–2000 1.0 × 10-1 pCi/L No threshold

U-234 (STRL–TARL) Urinalysis 2001–2009 1.0 × 10-1 pCi/L No threshold

U-234 (STRL–TARL) Fecalc 2001 7.1 × 10-2 pCi/24 hr No threshold

U-235 (Eberline) Urinalysis 1999–2000 1.0 × 10-1 pCi/L No threshold

U-235 (STRL–TARL) Urinalysis 2001–2009 1.0 × 10-1 pCi/L No threshold

U-235 (STRL–TARL) Fecalc 2001 7.3 × 10-2 pCi/24 hr No threshold

U-238 (Eberline) Urinalysis 1999–2000 1.0 × 10-1 pCi/L No threshold

U-238 (STRL–TARL) Urinalysis 2001–2009 1.0 × 10-1 pCi/L No threshold

U-238 (STRL–TARL) Fecalc 2001 7.1 × 10-2 pCi/24 hr No threshold

a. Values from review of available bioassay records. Values could be applicable earlier or later, but 

documentation was not found.

b. Reporting levels shown were not documented as the MDA but can be interpreted as the MDA in lieu of other 

information.
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 73 of 131

c. Values based on one fecal sample analysis in 2001.

d. Value below which “no exposure was assigned” from O’Connell (1975).

e. Value from Murray (1994).

f. Values for STRL – TARL are contractual required detection limits.

g. Values from employee bioassay data records.

h. Value from Scarpitta (2001).

i. Value from BNL (1958).

j. Value from Miltenberger (1982).

Table 5-4. Detection limits for in vivo bioassay.

Radionuclide

Method/

description Period MDA (nCi)a Reportinglevel (nCi)b

Am-241 WBC 2003–2009 2.33–9.47

Ba-133 WBC 2001 23.1

Ba-140 WBC 2001 22.1

Be-7 WBCc 1992–1999 25

WBC 1985–1987 45

WBC 1973–1974 5

WBC 2002–2009 25

WBC (dual detector) 1992–2001 50

WBC (single detector) 1993 82

Cd-109 WBC 2001 20.1

Ce-139 WBC 1999–2009 1.76–9.0

Ce-144 WBC 2003–2008 13.3–27.1

Co-57 WBCc 1993–1999 3.14–11.1

WBC 1986–2009 1.24–13.0

Co-58 WBCc 1992–1999 0.95–7.81

WBC 1973 5

WBC 1973–2009 2.39–5.0

Co-60 WBCc 1992–1999 1.19–8.13

WBC 1987 2.1

WBC 1986–2009 1.29–9.20

Cr-51 WBCc 1992–1998 14.8–79.1

WBC 2001 203

Cs-134 WBCc 1992–1999 0.851–5.41

WBC 1979 1

WBC 1987 2.4

WBC 1986–2009 2.09–10.3

Cs-137 WBCc 1992–1999 1.02–7.06

WBC 1987 3.4

WBC 1973–1980 1

WBC 1986–1988 3.0–6.0

WBC (dual detector) 1993–2009 2.64–6.08

WBC (single detector) 1993 8.5–9.2

Fe-59 WBCc 1993–1999 2.01–7.41

WBC 1986–1988 6.0–8.0

WBC 1999–2009 4.0–7.8

Hg-203 WBCc 1998 1.85–4.10

I-123 WBCc 1985–1998 4.64–8.0

I-125 Thyroid 1984 0.3

Thyroid 1979–1980 4.1–5.4

Thyroid 1984–1989 0.30–0.50

I-131 WBCc 1992–1998 2.19–6.64

Thyroid 1989 4

WBC 1961 10

WBC 1986 9

WBC 1987 4
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 74 of 131

Radionuclide

Method/

description Period MDA (nCi)a

Reporting

level (nCi)b

WBC 1973–1974 5

WBC 1988–1989 2.3–8.3

WBC 1999–2008 1.92–6.0

WBC (dual detector) 1993 3.8

WBC (dual detector) 1994 3.3

WBC (single detector) 1993 8.8

I-133 WBCc 1992–1998 0.85–11.0

WBC (dual detector) 1994 3.4

Mn-54 WBCc 1992–1999 0.966–8.21

WBC 1987 2.5

WBC 1979–1980 1

WBC 1986–1988 2.0–4.0

WBC 1999–2009 2.22–3.74

WBC (dual detector) 1993–1994 4.7

WBC (single detector) 1993–1994 7.5

Na-22 WBCc 1993–1999 1.03–8.87

WBC 1999–2009 2.58–4.03

Na-24 WBCc 1995–1996 1.29–5.18

Os-185 WBCc 1995–1998 1.03–4.87

Sc-46 WBCc 1997–1998 1.0–3.99

Sn-117m WBCc 1995–1998 4.53–6.06

Sr-85 WBCc 1999–2008 2.21–7.0

Tl-208 WBCc 1992–1993 3.60–9.21

U-235 Lung 1973–1979 0.2

WBC 1973 1

Xe-127 Lung 1985–1987 4.7

Y-88 WBC 1973 5

Zn-65 WBCc 1992–1999 2.29–15.6

WBC 1973–1974 2

WBC 1986–1988 7.0–8.0

WBC 1999–2008 5.0–9.18

WBC (dual detector) 1993–1994 8.50

WBC (single detector) 1993–1994 19.1

Zr-95/Nb-95 WBCc 1992–1994 4.53–16.4

WBCc 1995–1998 1.65–6.93

a. Range of values found in bioassay records and BNL correspondence. Individual records will 

usually include this information for the worker.

b. Reporting level was the MDA.

c. The bioassay reports may indicate a “lung geometry” was used, however, this refers to the calibration 

phantom and results should be treated as a whole body count.

5.9 EXCRETA SAMPLE KIT CODES

No codes have been found for excreta sampling kits. Routine samples for H-3 were grab samples 

(BNL 1993) to be collected weekly or within 4 hours of completion of a task that required bioassay. 

Twenty-four-hour samples could also be collected when other radionuclides were suspected (BNL 

1993). In 1984, 24-hour kits consisted of three 500-ml bottles (Miltenberger and Steimers (1984). In 

1999, spot samples were collected in 60-cm3 plastic bottles (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. 02 Effective Date: 02/07/2013 Page 75 of 131

generally not available. For some facilities, stack emissions or other measurements could be useful 

for dose reconstructions. Table 5-5 lists this information. In all cases, specific information about 

particle size was not found; the default of 5-μm activity median aerodynamic diameter is assumed to 

apply. In addition, specific information about solubility type was not found; the solubility that is most 

favorable to the claimant should be applied.

Table 5-5. Solubility type and fraction activity by facility.

Facilitya Compoundb Radionuclide

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

Buildings 701 to 

703, 704 

(fanhouse), 708 

(instrument house), 

709 (canal house), 

709A (canal water 

treatment facility) 

(1950–1957)

Natural uranium metal fuel, 1- × 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

Buildings 701-703 

(1958–1969)

Enriched uranium fuel, in the form of curved plates of 

uranium-aluminum 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 to 1969 environmental monitoring reports 

(Meinhold and Hull 1998).

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
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 76 of 131

Facilitya Compoundb Radionuclide

Activity 

fraction

BGRR

Buildings 701-703 

(1970–2006)

Residual contamination 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

BMRR

Buildings 490-491 

(1959–2000)

Enriched uranium fuel, in the form of curved plates of 

uranium-aluminum 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 laboratories 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 Building 

(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

Buildings 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 

Building 919

(1977)

Handling of tritium sources up to 100 mCi. 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)

Building 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

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
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Facilitya Compoundb Radionuclide

Activity 

fraction

Physics

T-248, T-109, others 

before 

(1947–1951)

Polonium used in physics experiments. Po-210

Physics

Building 510

(1962–present)

Processing irradiated targets, up to 1 mCi; 5,000 Ci of Co60 in the GIF (1977).

Radiation and 

Chemical 

Technology

Buildings 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.

Engineering Test 

Building 528

(1963–present)

Contained Co-60 sources up to 10,000 curies which were 

removed by the mid-1970s.

Low activity sources for instrument testing

Co-60

High Intensity 

Radiation 

Development 

Laboratory

Building 830

(1951–present)

Contained million-Ci Co-60 and Cs-137 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 GIF was still in 

operation (1970 – 1990s).

Fabrication of solid sources up to Ci amounts; 

Cs-137

Co-60

Hot Machine Shop

Building 530

(1947–1975)

Replaced by

Building 462

Machining radioactive solids (1977), including uranium and 

radioactive graphite. Repair work on “hot” fuel-handling 

tools.

Health Physics & 

Safety, 

Instrumentation

Building 535

(1964–present)

Radioactive solutions up to 100 μCi (1977).

Chemistry

Building 555

(1966–present)

Radioactive solutions and irradiated targets up to 1 mCi 

(1977).

Reclamation and 

Hot Laundry

Building 650

(1959–1996)

Radioactive decontamination up to mCi amounts (1977). 

82% of activity released to sewer in 1973 was H-3, 

remainder was gross beta (ERDA 1977, p. 125).

HFBR

Building 750

(1965–1997)

“Fully” enriched uranium and aluminum alloy, heavy-watercooled and -moderated reactor that operated up to 40 MWthermal (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

Building 901, 901A

(1970–present)

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, 

RARAFc

Building 902

(1968–present)

Irradiated target processing, up to 100 μCi (1977).
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Facilitya Compoundb Radionuclide

Activity 

fraction

BLIP Facility

Buildings 930 and 

931B

(1972–present)

Processing irradiated samples containing short-lived 

radionuclides (1977) for medical use; targets were sealed 

during normal operations but radioactive gases (N-13, O14, 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 1 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

Facilitya Compoundb Radionuclide

Activity 

fraction

Waste Management 

Incinerator

Building 445

(1981–present)

BNL granted a permit by New York in March 1981. 

Fractions based on averages from 1981 to 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. RARAF = Radiological Research Accelerator Facility.

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

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- and processspecific resuspension factors are not likely to be available. During the operation of the HFBR, the 

airborne tritium contamination was 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 spent 

4 hr/d on the operations level and 4 hr/d on the equipment level (BNL 1999b).

5.13 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 
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 79 of 131

occupational exposure (NIOSH 2002, p. 32). BNL was not a processing or storage location for large 

quantities of 226Ra or 222Rn [15]. Onsite 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.

5.14 GUIDANCE FOR DOSE RECONSTRUCTORS

During the evaluation of information for SEC determination, NIOSH determined that bioassay data 

after 1993 was available from the site. Therefore, if the energy employee has no bioassay data since 

1993, it should be assumed that the energy employee did not 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 for all years of employment. If there is no bioassay 

and no tritium doses reported with external doses, then no tritium dose should be assigned for any 

given year. If there is no tritium bioassay, but there is a tritium dose with the external report, dose 

reconstructors should use the larger of:

• The reported dose (lognormal distribution with a GSD of 3; or

• The missed dose (triangular distribution) 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 μCi would be 11 mrem (0, 0.011, and 0.022 would be the IREP input parameters).

6.0 OCCUPATIONAL EXTERNAL DOSIMETRY

6.1 INTRODUCTION

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 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. It 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 startup through 1984, at which time 

the Laboratory switched to a vendor (Lane and Reciniello 2004, Yoder 2005). 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.
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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 Early Dosimetry for High-Energy Proton Accelerators

Design and construction of the first GeV proton accelerators in the 1950s demanded an increased 

understanding of high-energy accelerator radiation environments. When a beam of protons in a highenergy (>500 MeV) accelerator strikes its target, a variety of charged and uncharged particles are 

produced, including neutrons, neutrinos, pions, kaons, muons, electrons, and positrons. The 

Cosmotron, which operated from 1952 to 1966, was initially only partially shielded. The shielding was 

improved with time as the beam current and available run times increased. For personnel dosimetry, 

the desired approach was to measure each component (neutron, photon, etc.) of the radiation dose 

and multiply by an appropriate relative biological effectiveness (RBE) value to account for the fact that 

some types of radiation cause a given effect at lower doses than do X-rays and gamma rays. 

However, initially only very limited information and measurement techniques were available. At BNL a 

method was devised in which NTA film was used as the primary dosimeter because the beta-gamma 

film was too insensitive for the weekly exchange cycle being used. Beta-gamma film was included in 

the badges because they might have been worn in other laboratory areas and would have been useful 

in case of an overexposure at the accelerator. The NTA film was calibrated with actual machine 

radiation and a tissue-equivalent ionization chamber. Because the calibration varied considerably 

with location, a conservative value of mrem per track on the film was used to interpret the results 

(Cowan 1969). Even though the NTA film responded primarily to neutrons, the calibration was to the 

total dose equivalent from all radiations. This was viable because neutrons were expected in all 

locations. NTA film dosimeters were required for all personnel at the facility. Administrative controls 

were applied to keep people away from the direct beam and areas with high levels of scattered 

radiation. At the Cosmotron, individuals were primarily exposed to scattered radiation that had an 

RBE in the range of 3 to 5, but an RBE of 10 was applied to personnel measurements. However, the 

weekly exchange resulted in poor statistics even for the NTA film (Cowan 1966). After the Cosmotron 

was fully shielded, conventional methods, using NTA film calibrated with Pu-Be neutrons and betagamma film calibrated with 60Co, were used. This method did not underestimate the dose equivalent 

in the areas where most exposures occurred and considerably overestimated them in other areas. 

Therefore, it was judged to be satisfactory for routine use (Cowan 1969).

The AGS, which began operations in 1960, was a well-shielded 30-GeV proton accelerator. In such 

machines, neutrons constitute the primary hazard outside of the shielded areas of the facility (NCRP 

2003). Initially the approach to personnel dosimetry used at the Cosmotron was applied. However, in 

the early 1960s a measurement technique was developed using a tissue-equivalent proportional 

counter to determine the distribution of linear energy transfer (LET) under specific measurement 

conditions. The term quality factor (QF), which was more broadly representative of the effects 

observed in biological studies, replaced RBE for radiation protection purposes. In 1954, values of QF 

as a function of LET had been published by the NCRP (NCRP 1971). Therefore, it became possible 

to determine the mean QF without a detailed knowledge of the radiation components. This served as 

a means of ensuring that the personnel doses as recorded on the NTA and beta-gamma films were 

providing conservative dose estimates. The term QF has since been replaced by the radiation 

weighting factors.

From the onset, BNL recognized the hazards of direct exposure to the beams of high-energy protons. 

Access to areas where the beams were brought out through the shielding, either scattered from a 

target or by ejection of a portion of the internal beam, had to be carefully controlled. Not only was it 

less certain that the values of QF being assumed would provide a sufficient safety factor but, in 

addition, exposure to a direct beam could result in a part of the body being irradiated that did not 

include the film badge (Cowan 1963, p. 14). Shortly after the AGS commenced operation, BNL found 
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that a considerable flux of μ-mesons (muons) was penetrating 16 ft of heavy concrete and emerging 

into the experimental area. They were then faced with determining whether dose due to this flux was 

being properly evaluated, particularly because high-energy muons are minimum ionizing particles and 

the NTA film used for personnel monitoring did not record their tracks. Muons decay into high-energy 

electrons plus neutrinos. The high-energy electrons interact with matter to produce high-energy 

bremsstrahlung radiation which, in turn, can produce photoneutrons. Alternately, the bremsstrahlung 

radiation can produce more electrons and positrons by pair production, and the resulting electronphoton cascade can extend through several generations depending on the initial electron energy. 

Because pair production is the more likely of these last two processes, the number of neutrons 

present was relatively small. Therefore, the radiation emerging from the concrete into the 

experimental area consisted of a mixture of muons, electrons, gamma rays, and some neutrons. Two 

BNL beta-gamma film and NTA badges were exposed near a tissue-equivalent chamber to test the 

response of their personal monitors to the muon-caused radiation. Evaluation of the film badges by 

the normal procedure, which by now included the beta-gamma film, convinced the BNL health physics 

personnel that the dose equivalents being assigned provided an adequate safety factor (Cowan 1963, 

p. 12).

It appears that the BNL health physics personnel were aware of the unique problems associated with 

the dosimetry for high-energy accelerators. When information was lacking, they endeavored to 

choose values of RBE or QF that would overestimate the dose equivalent. However, the missed 

doses for the weekly exchange period were high. The values of missed dose to be used in dose 

reconstructions are given in Section 6.5.

6.4.2 Neutron Dosimetry

The neutron doses were measured with NTA film from the beginning of operations through 1995. 

Particles from high-energy accelerators produced stars from spallation products on the NTA film, and 

doses were evaluated by counting the number of prongs on the stars. The total number of prongs 

were 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. An RBE and later a QF of 10 was used 

from the very beginning of operations of the accelerators and cyclotrons for these products (Cowan 

ca. 1953). Vendor-supplied NTA and CR-39 plastic track detectors were tested at the AGS starting in 

1986 (Gilmartin 1986f). Lexan was added to the badges for the AGS as of December 5, 1987, and 

CR-39 was used in the badges for the reactors and NSLS. All badges continued to include NTA film 

(Schopfer 1987) through 1995.

Due to problems with the Lexan dosimeter, its use was discontinued in 1995. BNL site experts report 

that the highest of credible dosimeter values were used as the doses of record during this period. 

Frequently this was the NTA film. Where vendor-supplied detailed records are available, it should be 

possible to identify which measurement is reported. Column (5) of the report shows which dosimeter 

was used to assign the neutron dose using the following codes: (P) NTA Film; (B) Neutrak 144 (CR39); and (E) Neutrak 1 (Lexan) (Kahnhauser 2011). After 1995 the neutron dosimeter for all facilities 

consisted of a multipurpose TLD badge. A special-purpose albedo TLD badge was also available. To 

monitor high-energy neutrons from accelerators, a CR-39 dosimeter in a plastic bag was attached to 

the multipurpose dosimeter (Sengupta 2000).

For a best estimate during the period of overlapping dosimeters (1987 to 1995), dose reconstructors 

should consider using the recorded value because it is based on the judgment made by the site 

experts at the time. As stated before, this is likely to be the highest measured dose, but it might or 

might not include the Lexan dose because that was at times considered to be unreliable. 

Overestimates could include a sum of all of the measured values that are available for each monitored 

period, using the assumption that the two (or possibly three) dosimeters responded independently to 

different energy ranges when, in fact, there was significant overlap.
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6.4.3 Special Dosimetry

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 measured radiations were recorded as WB dose or dose equivalent. The 

correct value for the RBE (later the QF) was determined to be conservative; because a value of 10 

was always used (Patterson and Thomas 1994; Cowan 1963).

6.4.4 Lost or Destroyed Dosimeters

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).

6.5 MISSED DOSE

Missed doses for monitored employees at BNL result 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 historical missed dose 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) 

were conservative and resulted 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

Exchange

frequency

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 30d Weekly

Monthly

780

180

1987–1995 Lexanc 30 Monthly 180

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 (2007).

c. Processing done by RS Landauer. In addition to NTA film, Lexan was added for the AGS and 

CR-39 was added for the reactors and NSLS from 1987 to 1995. Testing of these plastic

dosimeters began in 1986.

d. Based on the analysis of background tracks and historical calibration factors presented in 

Attachment A.

6.5.1 Unmonitored Worker

Unmonitored workers are those 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 
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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).

Doses for unmonitored employees could be as much as 10% of the relevant standard, which was 

1,500 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].

6.5.2 Annual Average Worker Dose

Radiation workers will likely have dosimetry information available in the files for a claim. For workers 

who were not considered to be radiation workers, but who occasionally entered facilities with radiation 

sources, an overestimate of external dose can be obtained from the annual average external dose 

values presented in this section.

Summary information is available on annual doses received by BNL workers and visitors from 1958 

through 2009. This information was gathered and summarized in the SEC Evaluation Report for the 

first BNL SEC (NIOSH 2009). The following discussion presents a summary of the information 

provided in that report and additional information used to determine average annual doses for 

employees and visitors of the Brookhaven National Laboratory. The dose values in the SEC 

Evaluation Report were supplemented for 1971 (Moser ca. 1972), 2008 (DOE 2010), and 2009 (DOE 

2011).

The dose values available are the annual average dose to BNL workers and visitors based only on 

the individuals with positive measured dose. Because only the individuals with positive measured 

dose are included, the average values represent a favorable to claimant estimate of dose for 

individuals who were not likely to be routinely exposed to external radiation fields but might have had 

intermittent potential for exposure. The term “visitor” as used here represents a non-Brookhaven 

employee and includes visiting scientists, construction workers, and others who visited the site that

were required to be monitored for entry into radiological facilities. The average dose was available for 

most years, except for 1958 through 1966, 1974, and 1975. For these years, the average available 

value represented an average over all employees and visitors. The employee and visitor doses were 

determined from the given overall average using ratios of doses for 1967 to 1973 and 1976 to 1979. 

For 1967 to 1970, only the employee dose was given. For these years the visitor dose was evaluated 

based on the binned dose values and the average measured employee dose. The binned data was 

represented by the number of measured badge readings within each dose bin of 0 to 1 rem, 1 to 2 

rem, 2 to 3 rem, 3 to 4 rem, 4 to 5 rem, 5 to 6 rem, 6 to 7 rem, 7 to 8 rem, and 8 to 9 rem. First, an 

estimate of the average dose was evaluated using the midpoints of the bins multiplied by the number 

of badges recorded in each bin, summed over all bins. This provided an overestimate of the average 

dose in comparison with the reported average values. The ratio of the reported average to the 

estimated average for the employee category was used to adjust the estimated average for the 

visitors to obtain an average for the visitors. 

For 2009 and 2010, the employee average was the only value available. The visitor average was 

estimated using the 2008 ratio of visitor-to-employee average dose. This provided a higher estimate 

of visitor dose in comparison with the ratio for other recent years.

No dosimetry information was available for 1947 through 1957. The values for 1958 represent a 

reasonable estimate for these years because the potential for external dose was less in the early 

years as facilities and programs were being developed.

In addition to the doses based on reported annual average values, the potential missed dose has also 

been evaluated based on guidance in ORAUT-OTIB-0020, Use of Coworker Dosimetry Data for 
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External Dose Assignment (ORAUT 2011c). A potential missed dose was evaluated for 11 zero 

badge readings per year (assuming a monthly badge exchange frequency) to maximize the potential 

external doses. The annual average dose values and the total dose (including the missed dose) are 

presented in Table 6-2. The values in this table can be used when the available information indicates 

the worker might have been exposed intermittently but there are no dosimetry records in the files. 

Because the records for radiation workers appear always to include external dosimetry information, it 

is unlikely a radiation worker would not have external dosimetry available. Because the values in 

Table 6-2 are annual average values they should be used only for unmonitored workers with an 

indication in their files that they intermittently entered radiation zones. The dose values should be 

assigned as a constant because they represent an overestimate for unmonitored workers.

Table 6-2. Annual average workplace external dose.

Yeara

Employees Visitors Year Employees Visitors

Measured Totalb Recorded Total Year Measured Total Recorded Total

1958 0.312 0.477 0.223 0.388 1985 0.215 0.380 0.107 0.272

1959 0.317 0.482 0.227 0.392 1986 0.224 0.389 0.086 0.251

1960 0.297 0.462 0.213 0.378 1987 0.138 0.303 0.056 0.221

1961 0.289 0.454 0.207 0.372 1988 0.145 0.310 0.051 0.216

1962 0.272 0.437 0.195 0.360 1989 0.115 0.280 0.040 0.205

1963 0.271 0.436 0.194 0.359 1990 0.099 0.264 0.047 0.212

1964 0.276 0.441 0.198 0.363 1991 0.115 0.280 0.048 0.213

1965 0.276 0.441 0.198 0.363 1992 0.098 0.263 0.029 0.194

1966 0.315 0.480 0.225 0.390 1993 0.108 0.273 0.039 0.204

1967 0.277 0.442 0.225 0.390 1994 0.150 0.315 0.124 0.289

1968 0.279 0.444 0.219 0.384 1995 0.175 0.340 0.121 0.286

1969 0.192 0.357 0.172 0.337 1996 0.129 0.184 0.033 0.088

1970 0.201 0.366 0.171 0.336 1997 0.086 0.141 0.018 0.073

1971b 0.353 0.518 0.280 0.445 1998 0.099 0.154 0.019 0.074

1972 0.385 0.550 0.279 0.444 1999 0.052 0.107 0.024 0.079

1973 0.412 0.577 0.280 0.445 2000 0.058 0.113 0.038 0.093

1974 0.428 0.593 0.306 0.471 2001 0.043 0.098 0.028 0.083

1975 0.455 0.620 0.325 0.490 2002 0.070 0.125 0.018 0.073

1976 0.363 0.528 0.186 0.351 2003 0.046 0.101 0.019 0.074

1977 0.387 0.552 0.245 0.410 2004 0.080 0.135 0.070 0.125

1978 0.271 0.436 0.148 0.313 2005 0.048 0.103 0.013 0.068

1979 0.231 0.396 0.149 0.314 2006 0.041 0.096 0.046 0.101

1980 0.107 0.272 0.023 0.188 2007 0.033 0.088 0.039 0.094

1981 0.185 0.350 0.129 0.294 2008 0.036 0.091 0.043 0.098

1982 0.230 0.395 0.136 0.301 2009 0.029 0.084 0.034 0.089

1983 0.280 0.445 0.122 0.287 2010 0.054 0.109 0.063 0.118

1984 0.253 0.418 0.135 0.300

a. For 1947 to 1957, use the 1958 values.

b. Total dose includes the annual average plus missed dose.

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. Table 6-3 lists the reactor and accelerator facilities and data

related to radiation types and energies. Table 6-4 provides this information for support facilities.

The photon radiation experienced by workers in the accelerator facilities occurs from radiation 

penetrating the shielding, which is designed primarily for stopping neutron radiation. This radiation 

tends to be high-energy photons that are able to penetrate the shielding material. A small component 

of Compton-scattered photons would also be expected in the radiation. Exposure to photons also 
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 85 of 131

occurs when workers enter the inner areas of the accelerator to perform maintenance, target 

changes, and experiment setup while the facility is shut down. This exposure results from activation 

products generated during operation. This photon energy is typically of high energy, >250 keV. 

Therefore, the photon energy range for accelerators that involve proton, neutron, and nuclei beams 

has been set as indicated in Table 6-3. This photon energy distribution is consistent with the 

distribution applied to accelerator facilities at other sites.
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 86 of 131

Table 6-3. Electron and photon energies and percentages for reactors and accelerators.

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

LMFR/820 Reactor 1957 1959

MRC490/490A Radioisotopes 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

10

90

PET/490 Accelerator 1977 Present Electron

Photon

>15

>250

100

100

Cosmotron/same Cyclotron 1952 1966 Photon 30–250

>250

10

RHIC/same Accelerator 1999 Present 90

AGS/same Synchrotron 1960 Present

Photon 30–250

>250

10

90

AGS experimental area/912 Accelerator 1960 Present

AGS ring/913 1960 Present

AGS booster/942 1992 Present

200 MeV LINAC/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

30–250

>250

100

10

90

NSLS/725 Electron storage 1982 Present Electron

Photon

>15

30–250

100

100

Chemistry Dept./490A Accelerator Electron

Photon

>15

30–250

>250

100

10

90

Table 6-4. Electron and photon energies and percentages for (support facilities).

a

Process/building Description

Operations Radiation 

type

Energy

Begin End (keV) Percentage

NSRLb Space Laboratory 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 1970

s

Early 

1990s

Chemistry Dept/555 & 490A Chemistry R&D 1966 Present

LEAF/555 Laser electron 

accelerator

1998 Present

Instrument Division/356 Material irradiation 1960 Present

S&EP/348 Instrument calibration 1959 Present

Biology/463 Biological R&D 1950 Present

Oceanographic 

Sciences/318

Oceanographic R&D

Radiation Sciences/

703, 703W

Health physics & 

environmental 

support

Nuclear Waste Mgt./830 Waste handling
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 87 of 131

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 1000 keV with wide-ranging percentages.

b. NSRL = National Air and Space Administration Space Radiation Laboratory.

6.7 NEUTRON RADIATIONS AND PERCENTAGES

Table 6-5 lists facilities with neutron radiations. They are reactors, accelerators, general laboratories, 

and medical facilities. The reactors are no longer in operation; the last was shut down in 2000.

Table 6-5. 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 1.91

>2 43 1.32

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 on which Table 6-5 is based. For 

Figure 6-1, measurements in the thermal column were normalized such that the flux in the bare 

thermal column was 7 × 108 n/cm2

/s. The thermal flux in the instrument tunnel was 2 × 109 n/cm2/s. 

The converter plate was a 235U plate used to convert the thermal beam into a fission spectrum.
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 88 of 131

Figure 6-1. Neutron spectrum in the thermal column with the converter plate (left) and in the 

instrument tunnel (right).

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 
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 89 of 131

 Figure 6-2. Neutron energy spectrum at the

 Cosmotron (Sanna and O’Brian 1963).

(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 and Story 1964; Strom 1982). Caruthers and Story (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 2004; 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 

was January 1, 2010 (ORAUT 2011d). Therefore, no adjustment in the recorded neutron dose (as 

described in the last column of Table 6-5) is necessary for records after 2009.

Three available studies of neutron energy spectra provide information for the 60” cyclotron (Xie 1986) 

and the AGS Experimental Area (Presig 1995 and Presig 1996). The neutron spectra provided in 

these studies were evaluated based on the five IREP input neutron energy ranges. The neutron 

energy distribution indicated in Table 6-5 for accelerator facilities was found to provide a favorable to 

claimant dose assignment compared to the neutron spectra given in the three studies. 

6.8 RECORDED DOSE PRACTICES

Recorded dose practices at BNL are given in Table 6-6 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).
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6.9 INTERPRETATION OF REPORTED DOSES

Early personnel doses were reported in millirep 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 

Table 6-6. Recorded dose practices.

Year Dosimeter measured quantities Compliance dose quantities

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),b mrem 

Skin = NP + P

WB = P + FN + tritium

1996–present Beta or nonpenetrating, mrem

Photon (P), mrem

Fast Neutron (FN), mrem 

Skin = NP + WB

WB = P + 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.

OCAS-IG-001, External Dose Reconstruction Implementation Guideline (NIOSH 2007). For the 

period from startup through 1995, the “Exposure (R) to Organ Dose (HT)” dose conversion factors 

should be applied to the recorded and missed dose. From 1996 and later years, the “Deep Dose 

Equivalent (Hp(10)) to Organ Dose (HT)” dose conversion factors should be applied.

Table 6-7 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 the 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).

Table 6-7. 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 

visitor 

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.
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6.10 ADJUSTMENTS TO RECORDED DOSE

The site used NTA film as the primary means to monitor workers for exposure to fast neutrons from 

the site startup through 1995. Two issues related to use of NTA film to monitored neutrons are 

discussed here: NTA neutron track fading and angular dependence.

6.10.1 NTA Track Fading

The problem of NTA neutron track fading was known at the BNL site for many years with the first 

study of the effect described in monthly reports for November and December of 1950 (BNL 1950). 

The effect was also reported in 1967 (BNL 1967), 1972 (Distenfeld and Klemish 1972), 1973 (Phillips, 

Swezey, and Gilmartin 1973), and 1975 (Philips 1975) using standard BNL NTA dosimeters. A 

summary of these studies is presented in Table 6-8.

Table 6-8. Summary of NTA fading studies at BNL.

Year reported Exposure Temperature (°F) Humidity (%) Loss/week (%)a

1950 Not specified Ambient Ambient 9b

1967c PuF4 Ambient Ambient 12

1972c PuBe 76 43 11.8

1972 PuBe 78 50 18.0

1972 PuBe 72 61 18.3

1972 PuBe 76 Desiccated 4

1972 PuBe 90 71 42

1973 PuBe Ambient Desiccated 6

1975 PuBe Ambient Desiccated 3–6

a. The weekly loss values are evaluated based on a power function of time.

b. Based on one-third of tracks lost in 1 month.

c. Data for the 1967 and 1972 studies are averaged over a 4-week period.

Results of the 1972 study indicate the loss rate is dependent on temperature and humidity. The 

majority of neutron exposures at BNL were associated with work at the AGS in well-ventilated aircraft 

type hanger buildings with inside conditions similar to those outside (Distenfeld and Klemish 1972 

p 6). The 1967 study indicated that during the summer months the temperature and humidity would 

be higher than other periods of the year and fading could be of concern, although, a 2-week wear 

period was considered to be permissible (BNL 1967). Desiccation of the film in badges enclosed in 

thin plastic sealers was being investigated at that time.

For the period from site startup through 1984, a fading correction factor of 18% per week provides a 

reasonable overestimate of fading for NTA film dosimeters on an annual basis. For a monthly wear 

period with 1 week for reading of the dosimeter, an average time for fading would be 3 weeks. This 

corresponds to a loss of about 55% of the tracks. The recorded neutron doses should be corrected 

by applying a multiplication factor of 1.81 (inverse of 55%).

For 1985 through 1995, the NTA film dosimeters were provided by and read by Landauer. The 

calibration methods used by Landauer involved exposure of calibration dosimeters at the same time 

the batch of worker dosimeters was prepared. The calibration dosimeters were also read at the same 

time as the worker dosimeters (Faust 2011). This method ensured that track fading effects were 

inherently included in the calibration process. Therefore, no correction for track fading is necessary 

for this period.

6.10.2 NTA Angular Dependence

Kathren, Prevo, and Block (1965) studied the response of NTA film dosimeters as a function of 

neutron incidence angle for neutrons of energy 0.87 to 6.19 MeV. The data presented in that paper 
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 92 of 131

are based on the response of the film expressed as “tracks per field dose” and indicate that for 

neutrons above about 2 MeV the film response is low. This implies that for NTA film dosimeters 

calibrated with the beam at 90 degrees to the badge, the calibration curve results in an underestimate 

of dose for fast neutrons if the worker is not exposed entirely in the AP geometry. For neutron 

energies above about 3.5 MeV, the response was a maximum at 90 degrees and lower at lower 

angles. For lower neutron energies, a correction is not needed because there was an over-response 

of the film for some angles. For fast neutrons, this corresponds to a correction factor of 1.33 applied 

to recorded doses. The calibration angle of 90 degrees corresponds to an exposure geometry of 

100% AP. However, the application of such a correction factor is only appropriate if the worker was 

likely exposed in a field of neutrons impacting the worker from many directions. The use of the 

correction factor would not be appropriate for workers who were exposed in an AP geometry because 

the calibration matches the exposure conditions. Most dose reconstructions use the DCFs for AP 

geometry because these DCFs are the highest except for a few organs (red bone marrow, bone 

surface, esophagus, lung, and remainder organs) where the PA DCFs are greater.

Calculations were performed to compare exposures in the AP geometry (without angular dependence 

correction) and in the rotational geometry (with angular dependence correction). The calculations 

were performed for the neutron energy range from 2 to 20 MeV. The results indicate that for a few 

organs, use of the corrected rotational geometry results in a higher estimate of recorded dose. Table 

6-9 indicates the ratio of corrected dose (based on rotational geometry) to the dose for AP geometry 

without correction. A value greater than 1 indicates the rotational geometry DCFs with angular 

correction would give a higher neutron dose for cases where the DCF values from OCAS-IG-001, 

External Dose Reconstruction Implementation Guideline (NIOSH 2007) are used. As discussed 

above, the angular correction factor is applicable only to the energies above 2 MeV and should not be 

applied to lower energy neutrons. The factors in Table 6-9 can be applied to the recorded neutron 

dose calculated using the AP DCFs for the appropriate external organ. Figure 6-3 is a logic diagram 

for application of the angular correction factors in Table 6-9. A favorable to claimant approach would 

be to assign the correction only to organs having a correction factor greater than 1. For application of 

the best estimate methods (ORAUT 2005), the factor must be applied as part of the Monte Carlo 

analysis and not as a correction factor for all geometries.

Table 6-9. Ratio of NTA film neutron dose for rotational 

geometry with angular correction to AP geometry.

External organ of reference Ratio: corrected rotational/AP

Bladder 0.84

Bone (red marrow) 1.31

Bone (surface) 1.28

Colon 0.96

Esophagus 1.13

Breast (female) 0.87

Liver 1.02

Lung 1.12

Gonads (female–ovaries) 1.06

Remainder organs 1.17

Skin 1.33

Stomach 0.86

Gonads (male–testes) 0.80

Thyroid 1.05

Lymphoid tissue 0.96

As an example of application of the NTA film correction factor, consider the neutron energy 

distribution for accelerator facilities from Table 6-5. Assume there are positive neutron doses reported 

in the dosimetry records for the worker prior to 1996. First, the neutron dose for each energy range is 

evaluated using the AP geometry dose factors for the organ of reference. The calculated dose in the 
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 93 of 131

0.1 – 2 MeV neutron energy range is assigned without application of an angular correction factor. If 

the geometry is not AP, the calculated dose is in the 2 – 20 MeV neutron energy range is multiplied by 

the correction factor from Table 6-9 for the reference organ. If the dose is being evaluated as an 

overestimate, then only angular correction factors greater than 1.00 need be applied.

6.11 VOLUNTARY NEUTRON RADIATION EXPERIMENTS

From the early 1970s to the mid-1980s studies were conducted in which voluntary subjects were 

exposed to neutron radiation to measure body content of various elements.

Figure 6-3. Logic diagram for application of NTA film angular correction factor.

Positive recorded neutron

dose before 1996? No

Yes

Calculate neutron dose using

AP Geometry DCF values

Is work exposure geometry

AP

No

Are any neutrons of energy >

2 MeV?

Yes

No

No angular correction 

necessary

Yes

From Table 6-9, get angular 

correction factor for organ

Is angular correction factor No

greater than 1.0?

Yes

Apply angular correcton to 

calculated AP dose
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 94 of 131

Prompt-gamma WB neutron activation analysis was performed as part of a research program to 

investigate the interrelationship between diet, nutrition, and body composition. An essential 

requirement in the study was the measurement of the protein muscle mass or specifically total body 

nitrogen. The activation analysis was performed by placing the subject on a cot that moved over a 

neutron source for 30 minutes. The subject was exposed twice, once in the prone position and once 

in the supine position. The neutron source was a 238PuBe neutron generator that produced fast 

neutrons of about 5 MeV. The patient was exposed to the fast neutron beam with a premoderator of 

deuterium oxide in a tank between the source and the patient. The neutrons reaching the subject 

were still considered to be fast neutrons (Vartsky, Ellis, and Cohn 1979). During the exposure period 

the entire body received a neutron irradiation that activated the body nitrogen and hydrogen for a 

short period. While the subject was being irradiated, the induced isotope gamma radiations were 

counted by appropriate detectors. The received dose was reported to be 95.4% (for a QF of 10) from 

fast neutrons and 4.6% from high-energy gamma radiation (Vartsky, Ellis, and Cohn 1979). The total 

reported dose for the procedure varied with time due to modifications and upgrades to the system. 

Dose assignment should be based on the dose to the employee from the medical records for the 

claim. If no dose is given, then the reported value of 26 mrem should be assigned (Vartsky et al. 

1979), using the above neutron/photon distribution.

Delayed-gamma WB neutron activation analysis was used to evaluate calcium content of the body. 

To perform the analysis, the subject was first counted in a WBC to determine the background 

radiation level in the body. The subject was then moved to the neutron activation facility and 

positioned on a moveable cot with the polyethylene moderator material placed in position around the 

subject. The cot was then positioned in the neutron irradiation facility and the subject was irradiated 

for 5 minutes. A PuBe neutron source was used for the irradiations. After irradiation, the subject was 

returned rapidly to the WBC to count the induced radioactivity produced. The site calculated a WB

dose of 0.277 rem, 92.6% (for a QF of 10) from fast neutrons and 7.4 percent from high-energy 

gamma radiation (Cohn et al. 1972). If a different dose is indicated in the medical records, the 

indicated dose should be used with the above neutron/photon distribution. If no dose is given, then 

the 0.277 value should be assigned using the above neutron/photon distribution.

The neutron doses should be assigned to the 2- to 20-MeV energy range with no correction for 

angular dependence because the beam irradiates the body in the AP and PA geometries rather than 

a rotational geometry. An ICRP Publication 60 correction factor of 1.32 should be applied to the 

neutron dose evaluation (ICRP 1982). The gamma dose component should be assigned to the 

>250-keV energy range.

The medical records for the energy employee should be reviewed for indications of participation in the 

voluntary neutron exposure studies. If information in the telephone interview record, DOL records, or 

DOE records indicates the employee participated in these studies but the actual exposures are not 

included in the records, the dose reconstructor should request the voluntary study medical records 

from the MRC at BNL.

6.12 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 
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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 

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%).

In 1985, BNL contracted with R. S. Landauer Company for film badge services. During this period the

DOE published criterion under the DOELAP for dosimeters (DOE 1986) that required 30% accuracy 

with 95% confidence based on use of four quarterly badges to evaluate the annual dose. The 

criterion applied to beta, photon, and neutron dosimeters. Table 6-10 summarizes the specifications 

from Landauer in 1988 (Schopfer 1988).

Table 6-10. Landauer gamma film, NTA film, and CR-39 specifications.

Radiation

category

Source

used

Range 

maximum

(mrem)

Mean

accuracya

(mrem)

Precision

(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.

A few reports of quality control checks on Landauer’s performance were found. 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 5,000 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).

Based on the above information, an uncertainty of 30% is assumed for the period from 1985 to 1995 

for all dosimeters.

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 the site. The 

Landauer-supplied CR-39 dosimeters were continued. Bias and uncertainty from 1996 on are 

expected to be consistent with DOELAP standards. To be granted DOELAP accreditation, BNL was 

required to demonstrate compliance with dosimeter system requirements. This compliance was 

demonstrated in performance testing of the BNL dosimetry system (Cummings 1997). The tests 

involved exposing sets of 15 dosimeters to various radiation fields including beta, photon, neutron, 
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 96 of 131

and some combinations of these. The tested dosimeters included Harshaw 8814 (TLD-600, -700), 

Harshaw 8806 (TLD-600, -700), and CR-39. Of the 360 dosimeter tests, 97.8% were within the 30% 

overall uncertainty requirement. The average uncertainty for each set of 15 dosimeters was under 5% 

including uncertainties due to source standardization, dosimeter position, and scattered radiation not 

stemming from the phantom, but excluding uncertainty due to the dose equivalent conversion factors 

and the photon component of the neutron irradiations (DOE 1986). The overall bias of the test was 

centered about 0.0, but some of the tests showed bias as high as about 14% for the Harshaw 8814 

dosimeters (overestimate of dose) to as low as about 20% for the CR-39 neutron dosimeter 

(underestimate of dose).

The results were similar for all types of radiations and tested dosimeters. Based on this information, 

an uncertainty of 30% is assumed for the period from 1996 to the present. Table 6-11 summarizes 

the uncertainty factors for BNL dosimetry.

Table 6-11. Uncertainty factors.

Dosimeter Uncertainty (%)a

Two-element film dosimeter (1947–1963) beta/gamma 40

Multi-element Film (1964–1984) beta/gamma 30

NTA Film (1954–1984) (2–14 MeV) neutron 30

NTA Film (1954–1984) (0.1–2 MeV) neutron 30

NTA Film (1985–1995) Landauer, beta/gamma/neutron 30

Harshaw 8814 (1996–present) beta/gamma/neutron 30

Harshaw 8806 (1996–present) beta/gamma/neutron 30

CR-39 (1996–1997) 30

a. Estimated variability of measured dose based on dosimeter technology as used 

in site radiation fields for long-term workers.

7.0 ATTRIBUTIONS AND ANNOTATIONS

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

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

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

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

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

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

[1] Potter, Eugene. Oak Ridge Associated Universities (ORAU) Team. 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. ORAU Team. Principal Medical Dosimetrist. December 2009.

Review of claim file records from BNL.

[3] Elyse Thomas. ORAU Team. Principal Medical Dosimetrist. December 2009.

Review of claim file records from BNL.

[4] Elyse Thomas. ORAU Team. Principal Medical Dosimetrist. December 2009.

Review of claim file records from BNL.
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 97 of 131

[5] Robert Morris. ORAU Team. Principal Health Physicist. August 2006.

Review of actual films at BNL.

[6] Robert Morris. ORAU Team. Principal Health Physicist. August 2006.

Review of actual films at BNL.

[7] Potter, Eugene. ORAU Team. 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. ORAU Team. 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. ORAU Team. 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. ORAU Team. 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. ORAU Team. Principal Health Physicist. November 2009.

The uncertainty range values quoted are based on professional judgment and experience.

[12] Potter, Eugene. ORAU Team. Principal Health Physicist. November 2009.

These conclusions are based on the review of a number of WB count results in individual 

dosimetry files.

[13] Potter, Eugene. ORAU Team. Principal Health Physicist. November 2009.

These conclusions are based on the review of a number of WB count results in individual 

dosimetry files, as well as other records and interviews.

[14] Potter, Eugene. ORAU Team. Principal Health Physicist. November 2009.

These conclusions are based on the review of a number of bioassay data packages and WB

count records.

[15] Potter, Eugene. ORAU Team. Principal Health Physicist. November 2009.

Based on a review of records and site interviews.

[16] Potter, Eugene. ORAU Team. Principal Health Physicist. November 2009.

This statement is based on the historical monitoring thresholds.
Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 98 of 131

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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, “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., 1947b, “Chest X-Rays,” memorandum to J. Knox, July 23. [SRDB Ref ID: 55276]

Sunderman, F. W., 1947c, “Clinical X-ray Equipment,” memorandum to P. M. Morse, Associated 

Universities, Brookhaven National Laboratory, Upton, New York, September 30. [SRDB Ref 

ID: 55278]

Taulbee, T. D., 2009, Measurement and Model Prediction of Proton-Recoil Track Length Distributions 

in NTA Film Dosimeters for Neutron Energy Spectroscopy and Retrospective Dose 

Assessment, Dissertation, University of Cincinnati, Cincinnati, Ohio, March. [SRDB Ref ID: 

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Memorial Institute, Richland, Washington, September. [SRDB Ref ID: 14914]

Van Der Karr, M. 1993, “Reducing Background Radiation Levels at the Igloo Area by Relocating 12”

Concrete Vaults and Reducing Skyshine,” Associated Universities, Brookhaven National 

Laboratory, Upton, New York, March 11. [SRDB Ref ID: 48492]

Vartsky, D., K. J. Ellis, and S. H. Cohn, 1979, “In Vivo Measurement of Body Nitrogen by Analysis of 

Prompt Gammas from Neutron Capture,” The Journal of Nuclear Medicine, volume 20, 

pp. 1158–1165.

Wolf, B. S., 1948, “Tentative Program for Control of Beryllium Hazard in AEC Contracts,”

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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.

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 specific amount of energy from ionizing radiation that is 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)

Exposure in air, measured in units of roentgens, at the point of entry into the body (without 

backscatter).

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. Also called film badge.
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filter

Material used (1) in a dosimeter to adjust radiation response to provide an improved tissue 

equivalent or dose response and (2) in an X-ray machine to selectively absorb photons from 

the beam to reduce unnecessary exposure of individuals or to improve radiographic quality.

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, 

which, when introduced in the path of a given beam of radiation, reduces 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.
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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 

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, neutron 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.
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natural uranium

Uranium as found in nature, approximately 99.27% 238U, 0.72% 235U, and 0.0054% 234U by 

mass. The specific activity of this mixture is 2.6 × 107 becquerel per kilogram (0.7 microcuries 

per gram). See uranium.

neutron

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

neutron film dosimeter

Film dosimeter with a nuclear track emulsion, type A, film packet.

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 that 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.
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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.

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 moderate to higher energy photons and neutrons 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.
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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.

probability of causation (POC)

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Program 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) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. 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.
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reactor elements

Fabricated fuel and target components inserted into reactors.

relative biological effectiveness

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 93 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 9.3 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.
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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.

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, 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.

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.
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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.
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The minimum detectable dose (MDD) or MDL for NTA film was a function of the total area of the film 

that were read (Taulbee 2009). The total area is the product of the area of the individual field-of-view 

(FOV), which varied with the microscope that was used, and the number of FOV read. At BNL, 24 or 

25 FOV were routinely read. The difference in read area is only about 4% and has been ignored for 

this analysis.

To determine the MDD it is necessary to know the distribution of the background tracks and the 

calibration factor (CF) in mrem/track. Some background NTA data were found for BNL. The data 

were summarized in the Health Physics Summary report for May 1955 (BNL 1955). Two technicians 

scanned 1,794 control films. Three groups of 25 FOV were scanned for each film. BNL observed that 

the data fit a Poisson distribution with a mean of 0.4 tracks. To verify this, this analysis performed a 

least-squares fit to the data and obtained a value of the mean (and variance) of 0.36. The R2 value 

was greater than 0.99. The observed and fitted data are reproduced in Table A-1.

Table A-1. Data from a 1955 background study and the Poisson fit to the 

data (mean = variance = 0.36).

No. of

tracks

Observed Fitted Poisson

Frequency Proportion Probability

Expected

frequency

Cumulative

probability

0 3,686 0.6847 0.69768 3,755.5917 0.6977

1 1,153 0.2142 0.25116 1,352.013 0.9488

2 362 0.0672 0.04521 243.3623 0.9941

3 115 0.0214 0.00543 29.2035 0.9995

4 40 0.0074 0.00049 2.6283 1.000

5 14 0.0026 0.00004 0.1892

6 8 0.0015 0 0.0114

7 4 0.0007 0 0.0006

8 1 0.002 0 0

At the time of the BNL (1955) analysis, BNL observed that there were excess tracks in the range of 3

to 8 in comparison with the Poisson fit. They attributed this to a low level of exposure to the control 

films. The observed and expected data from the Poisson fit are plotted in Figure A-1.

The detection limit (LC, also referred to as the “decision level” or “critical level”) is the net number of 

tracks for reaching a decision that an exposure has been detected. The value is calculated so that 

there is only a 5% chance that a result above the LC is part of the background distribution (i.e., a false 

positive). Taulbee (2009) discussed three methods of calculating a detection limit for NTA film. One 

of the three was the method that was used in the modern testing standards (NVLAP and DOELAP). 

However, this method requires more information than is generally available for historical data. The 

other two methods produced essentially the same result for the BNL data. The classic Currie formula 

as cited in Taulbee (2009) will be used in this analysis. The formula may be written as:

Lc=2.33 x standard deviation of the background=2.33 x (0.36)0.5=1.38 (A-1)

The Poisson distribution can only have integer values and, when films are evaluated, only whole 

numbers of tracks are counted. BNL counted at least three groups of 25 FOV for calibrations and for 

significant results, and averaged them (Phillips 1974). Therefore, it is possible that a fractional 

number of tracks could have been recorded. However, only one fractional calibration factor was 

found in the site documentation.

ATTACHMENT A

TECHNICAL BASIS FOR THE MINIMUM DETECTABLE DOSE 

FOR NEUTRON NTA FILM AT BNL

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Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 130 of 131

Figure A-1. Plot of observed and expected tracks for background control films.

Equation A-1 indicates that the detection limit should fall between one and two tracks. If one track 

was selected (i.e., 1.398 is rounded to the nearest whole number), the false positive rate would be 

greater than 5%, but only slightly. This is seen in the cumulative probability column of Table A-1. If a 

value of two tracks was selected as the detection limit, the false positive rate would be less than 1%. 

Given that the background rate was essentially zero, BNL would probably have recorded doses for 

one or more tracks per 25 fields. The actual procedures BNL used in counting have not been 

recovered. For purposes of this analysis, a value of 1.4 tracks was assumed. 

To interpret the detection levels in terms of dose, it is necessary to select an appropriate dose CF. 

The CF varies with both the time (month and year) of the calibration and the energy of the neutron 

spectrum. For this analysis, the calibration values for 1974, after the site had 25 years of experience 

in evaluating NTA film, were assumed to apply for calculating the MDDs.

BNL, like most sites, calibrated the various emulsion batches with an alpha-neutron source. PoBe 

was used as the calibration source at BNL until about 1959 when it was replaced by PuBe (BNL 

1959). These sources have similar average neutron energies (PoBe ≈ 4.2 MeV; PuBe ≈ 4.5 MeV). 

Differences were ignored in this analysis. Although the badges were calibrated to an intermediate 

energy spectrum, in the field the badges were potentially exposed to highly moderated neutrons from 

0

500

1000

1500

2000

2500

3000

3500

4000

0 1 2 3 4 5 6 7 8

Tracks

Number of Cases of Background Tracks per 25 

Fields in 1955 and Expected Cases for a Least 

Squares Fit Poisson Distribution

Observed Cases Expected Cases (mean = 0.36)

ATTACHMENT A

TECHNICAL BASIS FOR THE MINIMUM DETECTABLE DOSE 

FOR NEUTRON NTA FILM AT BNL

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Document No. ORAUT-TKBS-0048 Revision No. 02 Effective Date: 02/07/2013 Page 131 of 131

reactors and to high-energy neutrons from accelerators. Very little calibration data by energy is 

available. However, data were found for 1974 and appear in Table A-2 (Phillips 1974). This data 

reflects the fact that the lower energy neutrons near the energy threshold of the NTA film are more 

difficult to detect than those from the calibration spectrum. The difference is a factor of 2. The table 

also shows the calculated MDDs assuming an LC of 1.4 tracks. The 1-MeV value (28 mrem) is the 

most favorable and was rounded up to 30 mrem in Table 6-1 for use in dose reconstruction.

Table A-2. 1974 calibration factors applied to the calculated LC in tracks.

FOV

Area

(mm2)

CF

(mrem/track)

LC

(tracks)

MDD

(mrem)

Calibration data from

Phillips 1974 (Phillips 1974)

25 0.442 20 1.4 28 1 MeV calibration

25 0.442 10 1.4 14 4.45 MeV calibration (PuBe)

25 0.442 5 1.4 7 17 MeV calibration

ATTACHMENT A

TECHNICAL BASIS FOR THE MINIMUM DETECTABLE DOSE 

FOR NEUTRON NTA FILM AT BNL

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