Page 1 of 109

DOE Review Release 01/17/2019 

Rocky Flats Plant – Occupational External 

Dose

ORAUT-TKBS-0011-6 Rev. 03

Effective Date: 01/14/2019

Supersedes: Revision 02 PC-1 

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

Document Owner 

Approval: Signature on File

Mutty M. Sharfi, Document Owner

Approval Date: 01/09/2019

Concurrence: Signature on File

John M. Byrne, Objective 1 Manager

Concurrence Date: 01/08/2019

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 01/08/2019

Concurrence: Signature on File

Scott R. Siebert, Project Director Representative

Concurrence Date: 01/08/2019

Approval: Signature on File

James W. Neton, Associate Director for Science

Approval Date: 01/14/2019

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION. 

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0011-6 Revision No. 03 Effective Date: 01/14/2019 Page 2 of 109

 

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

01/20/2004 00 New Technical Basis Document for the Rocky Flats Site –

Occupational External Dose. First approved issue. Initiated by 

Robert Meyer.

02/08/2007 01 Approved Revision to incorporate NIOSH-requested material, 

responses to questions by dose reconstructors, responses to 

Advisory Board on Radiation and Worker Health review, and 

attribution of certain statements in document. Constitutes a total 

rewrite of document. Added Section 6.11, Attributions and 

Annotations. Worker Outreach comments pertaining to CT-0205 and 

CT-0209 have been addressed. Incorporates internal and NIOSH 

formal review comments. The Worker Outreach comments from the 

June 23, 2004, meeting of the United Steelworkers of America Local 

8031 and Rocky Flats Security Officers Local Union 1 are addressed 

as follows: Lead aprons and port covers, Section 6.5.5; Neutron 

Dose Reconstruction Project, Sections 6.3.5.2, 6.4.2, and 6.7.3.4; 

and exposure geometry and gloveboxes, Sections 6.5, 6.6.4, 6.7.4, 

and 6.8.4. This revision results in an increase in assigned dose and 

a PER is required. Training required: As determined by the Task 

Manager. Initiated by Robert Meyer.

08/14/2007 02 Approved Revision 02 initiated to capture Advisory Board comments 

associated with the June 2007 Advisory Board meeting. The 

following OTIBs have been incorporated into this document during 

this revision: ORAUT-OTIB-0027, ORAUT-OTIB-0050, and ORAUTOTIB-0058. Incorporates formal internal and NIOSH review 

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

Initiated by Mutty M. Sharfi.

10/20/2010 02 PC-1 Page change initiated to update references and NIOSH required 

language on pages 10 and 11 in Section 6.1. Corrected text 

inconsistencies to indicate that non-affected original neutron dose, in 

addition to NDRP neutron dose, should be used in the 

reconstruction. Original neutron dose, in addition to notional neutron 

dose, should not be used in dose reconstruction. These changes 

occurred in pages 48 and 49 in Section 6.7.3.3. Updated references 

throughout and on page 65 in the Reference Section. No sections 

were deleted. No changes occurred as a result of formal internal 

review. Incorporates formal NIOSH review comments. Training 

required: As determined by the Objective Manager. Initiated by 

Mutty M. Sharfi.

01/14/2019 03 Revision initiated to capture Advisory Board comments and items 

associated with SEC-00192. Incorporates formal internal and NIOSH 

review comments. Constitutes a total rewrite of the document. 

Training required: As determined by the Objective Manager. 

Initiated by Mutty M. Sharfi. 
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TABLE OF CONTENTS

SECTION TITLE PAGE

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

6.1 Introduction .............................................................................................................................. 9

6.1.1 Purpose...................................................................................................................... 10

6.1.2 Scope......................................................................................................................... 10

6.1.3 Special Exposure Cohort............................................................................................ 10

6.2 External Dosimetry Overview ................................................................................................. 11

6.3 Interpreting the External Dosimetry Record............................................................................ 14

6.3.1 Dosimetry Records Systems....................................................................................... 14

6.3.2 Observed Data Discrepancies .................................................................................... 14

6.3.2.1 Rounding ..................................................................................................... 14

6.3.2.2 Deep Dose Not Equal to Gamma Plus Neutron Doses................................. 15

6.3.2.2.1 Possible Algorithm Issue ........................................................... 15

6.3.2.2.2 Possible Manual Correction....................................................... 15

6.3.2.3 Shallow Dose Minus Neutron Dose is Less than Photon Dose..................... 15

6.3.3 Database Table-Specific Issues ................................................................................. 15

6.3.3.1 RHRST_ED_TLD_HISTORY ....................................................................... 15

6.3.3.1.1 1976 Records (Individual Employed After 1976)........................ 16

6.3.3.1.2 1976 Records (Individual Employed Before 1976)..................... 16

6.3.3.1.3 Post-1976 Records.................................................................... 16

6.3.3.2 RHRST_ED_TLD_DOS ............................................................................... 16

6.3.3.2.1 Dose History Hard-Copy File Contents...................................... 17

6.3.4 Interpretation of Dosimetry Data................................................................................. 19

6.3.5 Additional Data Available............................................................................................ 19

6.3.5.1 Rocky Flats Work History File ...................................................................... 19

6.3.5.2 Neutron Dose Reconstruction Project File.................................................... 19

6.3.5.3 Job Exposure Matrix .................................................................................... 19

6.4 Historical Administrative Practices ......................................................................................... 25

6.4.1 Badged Population ..................................................................................................... 25

6.4.2 Badge Exchange Frequency ...................................................................................... 26

6.4.3 Field-Specific Calibration Factors ............................................................................... 27

6.4.4 Minimum Reported Dose............................................................................................ 27

6.4.5 Recorded Dose Practices ........................................................................................... 28

6.5 Common Issues ..................................................................................................................... 28

6.5.1 Number of Zero Readings .......................................................................................... 28

6.5.2 Discrepancies............................................................................................................. 31

6.5.3 Missing Entry.............................................................................................................. 31

6.5.4 Exposure Geometry.................................................................................................... 31

6.5.5 Lead Aprons............................................................................................................... 33

6.5.6 Recycled Uranium ...................................................................................................... 34

6.5.7 Potential Elevated Background Subtraction................................................................ 35

6.5.8 Badge Reading Policy, 1969 to 1970.......................................................................... 36

6.5.9 Gammacell 220 Cobalt-60 Irradiator........................................................................... 37

6.6 Photon Dose .......................................................................................................................... 39
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6.6.1 Energy Groups ........................................................................................................... 39

6.6.1.1 Default Exposure Spectra ............................................................................ 39

6.6.1.2 Dosimeter-Indicated Photon Energy............................................................. 40

6.6.1.2.1 Pre-1960 ................................................................................... 40

6.6.1.2.2 1960 to 1970 ............................................................................. 41

6.6.1.2.3 1970 to Present......................................................................... 41

6.6.2 Calibration Factor ....................................................................................................... 42

6.6.2.1 Reported-Dose-to-Organ-Dose Conversion Factor Units ............................. 42

6.6.3 Missed Dose............................................................................................................... 42

6.6.3.1 Limit of Detection ......................................................................................... 42

6.6.3.2 Number of Zero Readings ............................................................................ 43

6.6.3.3 Determination of Missed Dose ..................................................................... 43

6.6.3.4 Unmonitored Energy Range......................................................................... 43

6.6.4 Geometry.................................................................................................................... 44

6.6.4.1 Angular Dependence ................................................................................... 44

6.6.4.2 Exposure Geometry ..................................................................................... 45

6.6.5 Uncertainty ................................................................................................................. 45

6.6.5.1 Film.............................................................................................................. 45

6.6.5.2 Thermoluminescent Dosimeter..................................................................... 46

6.6.5.2.1 Loose-Chip Thermoluminescent Dosimeters ............................. 46

6.6.5.2.2 Panasonic Thermoluminescent Dosimeter ................................ 46

6.7 Neutron Dose......................................................................................................................... 47

6.7.1 Energy Groups ........................................................................................................... 47

6.7.1.1 Exposure Spectra......................................................................................... 47

6.7.1.2 Reported Dose to Energy Groups ................................................................ 48

6.7.2 Calibration Factor ....................................................................................................... 49

6.7.2.1 Dosimeter-Specific Quality Factor Conversion ............................................. 49

6.7.2.2 Reported-Dose-to-Organ-Dose Conversion Factor Units ............................. 49

6.7.3 Missed Dose............................................................................................................... 49

6.7.3.1 Limit of Detection ......................................................................................... 49

6.7.3.2 Number of Zero Readings ............................................................................ 50

6.7.3.3 Neutron Dose Reconstruction Project .......................................................... 51

6.7.3.4 Default Neutron-to-Gamma Ratio................................................................. 53

6.7.4 Geometry.................................................................................................................... 53

6.7.4.1 Angular Dependence ................................................................................... 53

6.7.4.2 Exposure Geometry ..................................................................................... 54

6.7.5 Uncertainty ................................................................................................................. 54

6.7.5.1 Film.............................................................................................................. 54

6.7.5.2 Thermoluminescent Dosimeter..................................................................... 54

6.8 Electron Dose ........................................................................................................................ 55

6.8.1 Energy Groups ........................................................................................................... 56

6.8.1.1 Exposure Spectra......................................................................................... 56

6.8.1.2 Reported Dose to Energy Groups ................................................................ 56

6.8.2 Calibration Factor ....................................................................................................... 57

6.8.2.1 Reported-Dose-to-Organ-Dose Conversion Factor Units ............................. 57

6.8.3 Missed Dose............................................................................................................... 57

6.8.3.1 Limit of Detection ......................................................................................... 57

6.8.3.2 Number of Zero Readings ............................................................................ 58

6.8.3.3 Unmonitored Energy Range......................................................................... 58

6.8.4 Geometry.................................................................................................................... 58

6.8.4.1 Angular Dependence ................................................................................... 58
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6.8.4.2 Exposure Geometry ..................................................................................... 59

6.8.5 Uncertainty ................................................................................................................. 59

6.8.5.1 Film.............................................................................................................. 59

6.8.5.2 Thermoluminescent Dosimeter..................................................................... 59

6.8.5.2.1 Loose-Chip Thermoluminescent Dosimeters ............................. 59

6.8.5.2.2 Panasonic TLD Dosimeter......................................................... 60

6.8.6 Skin Contamination .................................................................................................... 60

6.9 Unmonitored Individuals......................................................................................................... 62

6.9.1 In Production Areas .................................................................................................... 62

6.9.2 Outside Production Areas........................................................................................... 62

6.10 Extremity Dosimetry............................................................................................................... 62

6.11 Attributions and Annotations .................................................................................................. 63

References ........................................................................................................................................ 64

Glossary ............................................................................................................................................ 70

ATTACHMENT A EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS ..................... 75

ATTACHMENT B MAJOR JOB CATEGORIES ........................................................................... 91

ATTACHMENT C EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS 

PLANT ............................................................................................................93

LIST OF TABLES

TABLE TITLE PAGE

6-1 Interpretation of reported data................................................................................................ 20

6-2 Conservatively determined default dosimeter exchange frequencies ..................................... 29

6-3 Summary of historical recorded dose practices ...................................................................... 30

6-4 Dose limits based on exchange frequency ............................................................................. 32

6-5 Exposure geometry calculation .............................................................................................. 33

6-6 Exposure geometry defaults for major job categories............................................................. 33

6-7 Bias correction factors for application to dose received while wearing a lead apron............... 34

6-8 Photon energy distributions.................................................................................................... 39

6-9 Default photon energy distributions ........................................................................................ 40

6-10 Photon dose units for use with organ dose conversion factors ............................................... 42

6-11 Photon LODs .........................................................................................................................43

6-12 Uncertainty for photon film dose............................................................................................. 45

6-13 Uncertainty for loose-chip TLD photon dose........................................................................... 46

6-14 Uncertainty for DOELAP-accredited TLD photon dose........................................................... 47

6-15 Neutron dose measurements divided into energy groups....................................................... 48

6-16 Default neutron energy distribution......................................................................................... 48

6-17 Neutron dose units for use with organ dose conversion factors.............................................. 49

6-18 Neutron film track counting detail ........................................................................................... 50

6-19 Neutron LODs........................................................................................................................51

6-20 RFP lognormal neutron-to-photon ratio values, 1970 to 1976 ................................................ 54

6-21 ORAU Team-developed neutron-to-gamma ratios ................................................................. 54
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6-22 Uncertainty for TLD neutron dose measurements .................................................................. 55

6-23 Beta LODs .............................................................................................................................58

6-24 Uncertainty for beta film readings........................................................................................... 59

6-25 Uncertainty for loose-chip TLD beta dose .............................................................................. 60

6-26 Uncertainty for DOELAP-accredited TLD beta dose............................................................... 60

6-27 DU mixtures ...........................................................................................................................61

6-28 Activity per gram, 1-year-old DU ............................................................................................ 61

C-1 Missed external doses ........................................................................................................... 96

C-2 Lognormal neutron-to-photon ratio values, 1970 to 1976 ....................................................... 97

C-3 Annual external coworker doses for plutonium workers, 1952 to 1970 ................................. 101

C-4 Annual external coworker doses for plutonium workers, 1971 to 2005 ................................. 103

C-5 Annual external coworker doses for uranium workers, 1952 to 2005.................................... 104

C-6 Annual external coworker doses for plutonium workers, 1952 to 1970, modified in 

accordance with ORAUT-OTIB-0052 ................................................................................... 105

C-7 Annual external coworker doses for plutonium workers, 1971 to 2005, modified in 

accordance with ORAUT-OTIB-0052 ................................................................................... 107

C-8 Annual external coworker doses for plutonium workers, 1952 to 2005, modified in 

accordance with ORAUT-OTIB-0052 ................................................................................... 108

LIST OF FIGURES

FIGURE TITLE PAGE

6-1 External dosimeter history...................................................................................................... 12

6-2 ORNL-style film badge ........................................................................................................... 13

6-3 RFP multielement film badge ................................................................................................. 13

6-4 RFP interim TLD/film badge................................................................................................... 13

6-5 RFP Harshaw badge..............................................................................................................13

6-6 RFP Panasonic dosimeter...................................................................................................... 13

6-7 Portion of plant population badged......................................................................................... 25

6-8 Average measured worker dose............................................................................................. 26

6-9 Estimated beta dose rate from uranium metal at various enrichment levels ........................... 56

6-10 Shallow dose rate from natural uranium slab.......................................................................... 57

A-1 Occupational Dose Report reviewed June 4, 2002................................................................. 76

A-2 Occupational Dose Report reviewed June 4, 2002................................................................. 77

A-3 Dosimetry History by Individual query report dated March 10, 2003....................................... 78

A-4 Dosimetry History by Individual query report dated March 10, 2003....................................... 79

A-5 Dosimetry History by Individual query report dated March 10, 2003....................................... 80

A-6 Health Physics Yearly External Exposure Activity Run, 1953 to 1958 .................................... 81

A-7 Health Physics Yearly External Exposure Run, 1959 to 1963 ................................................ 82

A-8 Health Physics Yearly External Exposure Activity Run, 1964................................................. 83

A-9 External Dosimetry Detail, 1981 to 1983 ................................................................................ 84

A-10 Health Physics External Radiation Exposure Report, 1967 .................................................... 85

A-11 Radiation Dosimetry Individual Lifetime Report dated June 4, 2002....................................... 86

A-12 Radiation Health Records System – TLD Data....................................................................... 87

A-13 Radiation Dosimetry Detail Report, Termination Report dated September 17, 1996 .............. 88

A-14 Radiation Dosimetry Detail Report, Termination Report dated September 17, 1996 .............. 89

A-15 Occupational Radiation Exposure Information, 1988.............................................................. 90
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ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AP anterior-posterior

AWE Atomic Weapons Employer

Bq becquerel

Br brass (filter)

Cd cadmium (filter)

Ci curie

CY calendar year

d day

D&D decontamination and decommissioning

DDE deep dose equivalent

DOE U.S. Department of Energy

DOELAP DOE Laboratory Accreditation Program

DOL U.S. Department of Labor

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

EU enriched uranium

ft foot

g gram

GM geometric mean

GSD geometric standard deviation

H*(d) ambient dose equivalent at depth d in millimeters in tissue

HEU highly enriched uranium

HIS-20 Health Physics Information System (database)

Hp(d) personal dose equivalent at depth d in millimeters in tissue

Hp,slab(d) personal dose equivalent (slab phantom) at depth d in millimeters in tissue

hr hour

HSDB Health Sciences Database

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

IREP Interactive RadioEpidemiological Program

ISO isotropic

keV kiloelectron-volt, 1,000 electron-volts

LANL Los Alamos National Laboratory

lb pound

LOD limit of detection

MeV megaelectron-volt, 1 million electron-volts

mm millimeter

mo month

mR milliroentgen
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mrad millirad

mrem millirem

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

NDRP Neutron Dose Reconstruction Project

NIOSH National Institute for Occupational Safety and Health

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

NTA nuclear track emulsion, type A

ORNL Oak Ridge National Laboratory

ORAU Oak Ridge Associated Universities

OTIB ORAU Team technical information bulletin

OW open window

PER program evaluation report

PNAD personal nuclear accident dosimeter

PNL Pacific Northwest Laboratory

POC probability of causation

qtr quarter

R roentgen

RFP Rocky Flats Plant

RHRS Radiological Health Records System

ROT rotational

SEC Special Exposure Cohort

SOE Stationary Operating Engineer

SRDB Ref ID Site Research Database Reference Identification (number)

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

wR radiation weighting factor

yr year

μCi microcurie

§ section or sections
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6.1 INTRODUCTION

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

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

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

dose reconstructions at particular Department of Energy (DOE) or Atomic Weapons Employer (AWE) 

facilities or categories of DOE or AWE facilities. They will be revised in the event additional relevant 

information is obtained about the affected DOE or AWE facility(ies). These documents may be used 

to assist NIOSH staff in the evaluation of Special Exposure Cohort (SEC) petitions and the completion 

of the individual work required for each dose reconstruction.

In this document the word “facility” is used to refer to an area, building, or group of buildings that 

served a specific purpose at a DOE or AWE facility. It does not mean nor should it be equated to an 

“AWE facility” or a “DOE facility.” The terms AWE and DOE facility are defined in sections 7384l(5) 

and (12) of the Energy Employees Occupational Illness Compensation Program Act of 2000 

(EEOICPA), respectively. An AWE facility means “a facility, owned by an atomic weapons employer, 

that is or was used to process or produce, for use by the United States, material that emitted radiation 

and was used in the production of an atomic weapon, excluding uranium mining or milling.” 42 U.S.C. 

§ 7384l(5). On the other hand, a DOE facility is defined 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 [DOE] (except for buildings, 

structures, premises, grounds, or operations ... pertaining to the Naval Nuclear Propulsion Program);” 

and with regard to which DOE has or had a proprietary interest, or “entered into a contract with an 

entity to provide management and operation, management and integration, environmental 

remediation services, construction, or maintenance services.” 42 U.S.C. § 7384l(12). The Department 

of Energy (DOE) determines whether a site meets the statutory definition of an AWE facility and the 

Department of Labor (DOL) determines if a site is a DOE facility and, if it is, designates it as such.

Accordingly, a Part B claim for benefits must be based on an energy employee’s eligible employment 

and occupational radiation exposure at a DOE or AWE facility during the facility’s designated time 

period and location (i.e., covered employee). After DOL determines that a claim meets the eligibility 

requirements under EEOICPA, DOL transmits the claim to NIOSH for a dose reconstruction. 

EEOICPA provides, among other things, guidance on eligible employment and the types of radiation 

exposure to be included in an individual dose reconstruction. Under EEOICPA, eligible employment 

at a DOE facility includes individuals who are or were employed by DOE and its predecessor 

agencies, as well as their contractors and subcontractors at the facility. Unlike the abovementioned 

statutory provisions on DOE facility definitions that contain specific descriptions or exclusions on 

facility designation, the statutory provision governing types of exposure to be included in dose 

reconstructions for DOE covered employees only requires that such exposures be incurred in the 

performance of duty. As such, NIOSH broadly construes radiation exposures incurred in the 

performance of duty to include all radiation exposures received as a condition of employment at 

covered DOE facilities in its dose reconstructions for covered employees. For covered employees at 

DOE facilities, individual dose reconstructions may also include radiation exposures related to the 

Naval Nuclear Propulsion Program at DOE facilities, if applicable. No efforts are made to determine 

the eligibility of any fraction of total measured exposure for inclusion in dose reconstruction.

NIOSH does not consider the following types of exposure as those incurred in the performance of 

duty as a condition of employment at a DOE facility. Therefore these exposures are not included in 

dose reconstructions for covered employees (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
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6.1.1 Purpose

This technical basis document (TBD) is one part of the overall Rocky Flats Plant (RFP) site profile. 

The site profile describes plant facilities and processes, historical information related to occupational 

internal and external doses, and environmental data for use if individual worker recorded doses are 

unavailable. This document contains Part 6, Occupational External Dosimetry, of the RFP site profile. 

It provides necessary background information and critical data for the dose reconstructor to perform

individual dose reconstructions. Dose reconstructors will use this information as needed to evaluate 

external occupational doses for EEOICPA claims.

6.1.2 Scope

RFP operations played an important role in the U.S. nuclear weapons program. These operations 

included production of fissionable weapons components and waste management. This TBD contains 

supporting documentation to assist in the evaluation of occupational external doses from these 

processes using the methodology in OCAS-IG-001, External Dose Reconstruction Implementation 

Guideline (NIOSH 2007).

The methods and concepts of measuring occupational external doses to workers have evolved since 

the beginning of RFP operations. An objective of this document is to provide supporting technical 

data to evaluate, with assumptions favorable to claimants, the external RFP occupational doses that 

can reasonably be associated with worker radiation exposures covered under EEOICPA legislation. 

These doses include occupational external exposures in RFP facilities and onsite exposures to RFP 

environmental releases. This document addresses the evaluation of unmonitored and monitored 

worker exposure and missed dose. Consistent with NIOSH (2007), this document identifies how to 

adjust the historical occupational external recorded dose to account for current scientific methods and 

protection factors. 

Only a limited assessment of neutron doses can be performed before 1970. Unmonitored and 

notional neutron doses from 1952 through 1966 cannot be reconstructed under the EEOICPA. 

Between 1967 and 1970, unmonitored and notional neutron doses should be replaced with external 

coworker doses. Reported Neutron Dose Reconstruction Project (NDRP) and unaffected original 

neutron dose can be used for all years.

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

Modules for Bioassay Analysis (IMBA) computer code is a tool useful for internal dose calculations. 

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

document describes how to evaluate uncertainty associated with RFP exposure and dosimetry 

records. 

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

justification, or clarification of the associated information, are presented in Section 6.11.

6.1.3 Special Exposure Cohort

The Secretary of the U.S. Department of Health and Human Services has designated the following 

class of RFP workers as an addition to the SEC (Sebelius 2013): 

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

contractors and subcontractors who worked at the Rocky Flats Plant in Golden, 

Colorado, from April 1, 1952 through December 31, 1983, for a number of work days 

aggregating at least 250 work days, occurring either solely under this employment or in 
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combination with work days within the parameters established for one or more other 

classes of employees included in the Special Exposure Cohort.

NIOSH has determined that doses to unmonitored RFP workers from neptunium, thorium, and 233U 

(and its associated 232U and 228Th progeny) cannot be reconstructed from April 1, 1952, through 

December 31, 1983, inclusive (NIOSH 2013). The class includes all workers during the SEC period.

Based on the inability to reconstruct unmonitored doses from 1952 through 1983, as described above, 

all dose reconstructions for monitored workers during the SEC period are considered partial dose 

reconstructions. If monitoring data are available for workers in the SEC, dose is to be assigned as 

appropriate based on that data. Unmonitored neutron doses before 1967 have been identified as 

infeasible and therefore cannot be bounded. It is not feasible to reconstruct unmonitored neutron 

doses, in a bounding manner, before 1967; therefore, this TBD does not provide dose reconstruction 

guidance for unmonitored neutron doses before 1967. For all other years, external dose records 

and/or an unmonitored external dose reconstruction approach are provided. However, such dose 

reconstructions are still considered partial dose reconstructions because of the determination that 

exposure to neptunium, thorium, and 233U (and its associated 232U and 228Th progeny) during the SEC 

period cannot be bounded.

6.2 EXTERNAL DOSIMETRY OVERVIEW

Over the years, RFP used a variety of dosimeters to measure occupational ionizing radiation dose 

(Figure 6-1). Between 1951 and 1959, the Plant used a stainless-steel film badge based on an Oak 

Ridge National Laboratory (ORNL) design (Baker 2002). This was a two-element film badge with an 

open window (OW) and a 1-mm cadmium (Cd) filter. For the plutonium areas, in 1960, a brass (Br) 

filter with half the filtration of the cadmium filter was added to cover half of the OW. This provided 

separation of the 60-keV photons from the lower energy component. Very little information has been 

found on the performance of this dosimeter (Figure 6-2).

In 1964, a plastic film badge was introduced at RFP that included additional filters. In addition to the 

photon dosimetry system, this badge contained a personal nuclear accident dosimeter (PNAD; 

Figure 6-3). This portion of the badge was not used for routine personnel dosimetry (Baker 2002).

In 1969, a combination film and thermoluminescent dosimeter (TLD) badge was introduced at RFP, 

using TLD chips to measure photon dose. There were three TLDs in the lower part of the badge, 

covered with the same brass filter (two chips) and a thin cover (one chip) providing an OW. Film was 

used for neutron dose measurement. This badge contained a PNAD and was an interim badge 

(Figure 6-4) until the introduction of the TLD neutron system (Baker 2002).

In 1971, a full TLD badge was introduced at RFP that used TLD chips manufactured by the Harshaw 

Chemical Company (Figure 6-5). Referred to as the “Harshaw badge,” it contained a four-chip albedo 

neutron dosimeter (Falk 1971). Although the dosimeter did have a location for including a neutron 

film, this feature was not used. Photon measurement used three filter-covered TLDs, similar to those 

in the previous badge. This badge contained a PNAD.

In 1983, an automated Panasonic dosimetry system was introduced at RFP. This badge contained

two Panasonic dosimeters, one for measuring photon and beta dose and one for measuring neutron 

dose. The beta/photon dosimeter contained two TLD phosphors and a lead filter over one of the 

elements. The neutron dosimeter contained three neutron-sensitive elements and one neutroninsensitive element under cadmium or tin filters. This badge included a PNAD (Baker 2002).
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Figure 6-1. External dosimeter history. 

Year Holder

Beta/gamma

Neutron Extremity

Detector

Filtration

Deep Shallow Processor Detector Processor Holder Detector

1951 SS ORNL Std. X-ray 1-mm 

Cadmium

None LANL Track plate LANL SS ORNL Std. X-ray

1952 │ │ │ │ │ │ │ │ │

1953 │ │ │ │ │ │ │ │ │

1954 │ │ │ │ │ │ │ │ │

1955 │ │ │ │ │ │ │ │ │

1956 │ │ │ │ HPS │ HPS │ │

1957 │ │ │ │ │ NTA film │ │ │

1958 │ │ │ │ Rocky Flats │ Rocky Flats │ │

1959 │ │ │ │ │ │ │ │ │

1960 │ │ 1/2 Brass │ │ │ │ 1/2 Brass │

1961 │ │ │ │ │ │ │ │ │

1962 │ │ │ │ │ │ │ │ │

1963 │ │ │ │ │ │ │ │ │

1964 Plastic │ Multiple Multiple │ │ │ │ │

1965 │ │ │ │ │ │ │ │ │

1966 │ │ │ │ │ │ │ │ │

1967 │ │ │ │ │ │ │ │ │

1968 │ │ │ │ │ │ │ │ │

1969 Interim plastic TLD-700 │ │ │ │ │ │ │

1970 │ │ │ │ │ │ │ │ │

1971 Harshaw │ │ │ │ TLD-600/700 │ Harshaw TLD-600/700

1972 │ │ │ │ │ │ │ │ │

1973 │ │ │ │ │ │ │ │ │

1974 │ │ │ │ │ │ │ │ │

1975 │ │ │ │ │ │ │ │ │

1976 │ │ │ │ │ │ │ │ │

1977 │ │ │ │ │ │ │ │ │

1978 │ │ │ │ │ │ │ │ │

1979 │ │ │ │ │ │ │ │ │

1980 │ │ │ │ │ │ │ │ │

1981 │ │ │ │ │ │ │ │ │

1982 │ │ │ │ │ │ │ │ │

1983 Panasonic UD-802 │ │ │ UD-809 │ │ │

1984 │ │ │ │ │ │ │ │ │

1985 │ │ │ │ │ │ │ │ │

1986 │ │ │ │ │ │ │ │ │

1987 │ │ │ │ │ │ │ │ │

1988 │ │ │ │ │ │ │ │ │

1989 │ │ │ │ │ │ │ │ │

1990 │ │ │ │ │ │ │ │ │

1991 │ │ │ DOELAP │ │ │ Panasonic UD-813AS11

1992 │ │ │ │ │ │ │ │ │

1993 │ │ │ │ │ │ │ │ │

1994 │ │ │ │ │ │ │ │ │

1995 │ │ │ │ │ │ │ │ │

1996 │ │ │ │ │ │ │ │ │

1997 │ │ │ │ │ │ │ │ │

1998 │ │ │ │ │ │ │ │ │

1999 │ │ │ │ │ │ │ │ DOELAP │

2000 │ │ │ │ │ │ │ │ │

2001 │ │ │ │ │ │ │ │ │

2002 │ │ │ │ │ │ │ │ │

2003 │ │ │ │ │ │ │ │ │

2004 │ │ │ │ │ │ │ │ │

2005 InLight OSL │ │ Landauer CR-39 Landauer Luxel OSL

HPS = Health Physics Services; LANL = Los Alamos National Laboratory; NTA = nuclear track emulsion, type A; OSL = optically stimulated luminescence 

dosimeter; SS = stainless steel. Dates are approximate, overlap occurred during changeovers (Baker 2002). 1960: Brass not used on beta OW; no brass on 

wrist side of extremity holder (Baker 2002).
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Figure 6-2. ORNL-style film 

badge (including brass 

filter). 

Figure 6-3. RFP multielement film 

badge. 

Figure 6-4. RFP interim 

TLD/film badge. 

Figure 6-5. RFP Harshaw badge. Figure 6-6. RFP Panasonic dosimeter.
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6.3 INTERPRETING THE EXTERNAL DOSIMETRY RECORD

External dosimetry results are reported as:

• Penetrating (Pen) or deep (deep dose + neutron), 

• Skin (shallow dose + neutron), 

• Forearm (measured or estimated), and

• Hand (estimated).

The penetrating or deep dose is reported as the sum of the deep gamma and the neutron dose. The 

skin dose is reported as the deep dose unless the low-energy detector on the dosimetry badge 

indicated a response greater than the deep dose, in which case shallow gamma plus neutron were 

reported (Falk 1976). RFP did not use finger rings; however, wrist badges were used to assess the 

dose to the forearm. Hand dose was assessed using the dose measured by the wrist badge, and the 

application of a hand-to-wrist ratio (see Section 6.10).

6.3.1 Dosimetry Records Systems

In the 1950s, external dosimetry data were handwritten and reported manually. In the 1960s and 

early 1970s, information was maintained on early computer systems. The detailed data have not 

been carried forward. For the early years, the dose detail has been lost, and only quarterly totals are 

available. As noted, RFP typically summed the deep gamma dose and the neutron dose into a 

penetrating value. In the early years, the neutron and deep gamma numbers were not retained and 

only the penetrating value remains.

Electronic systems for which detailed data have been maintained include:

• HSDB (Health Sciences Database), 1976 to 1990; 

• RHRS (Radiological Health Records System), 1990 to 1999; and

• HIS-20 (Health Physics Information System, Canberra Industries), 1999 to 2006. 

In general, data migrated from one system to another. Little is known, or at least documented, about 

the precise method and decisions made during the migration of the HSDB data to the RHRS. 

However, the result of examining the contents of the data tables and hard-copy reports can be 

described.

6.3.2 Observed Data Discrepancies

The observations in Sections 6.3.2 and 6.3.3 are the result of an examination of available RFP 

dosimetry records (Savitz undated). 

6.3.2.1 Rounding

The electronic data in RHRS and many of the reports contain both gamma and neutron components 

as well as a deep dose equivalent (DDE). A manageable problem is exhibited by the rounding of 

individual deep dose values as well as the yearly or quarterly totals. It appears that rounding to the 

nearest millirem value occurred on the external deep dose after the values were added to calculate 

the DDE. In many cases, this results in a discrepancy of 1 mrem per measurement on the report 

cards. Depending on the exchange frequency for a particular worker, there could be a difference of 

several millirem.
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6.3.2.2 Deep Dose Not Equal to Gamma Plus Neutron Doses

In this case, the problem is clearly not due to rounding but rather to a discrepancy between the deep 

dose components and the deep dose value that is stored separately. The magnitude of the 

discrepancy is greater than 1 mrem. Two specific situations have been identified, as described in the 

following sections.

6.3.2.2.1 Possible Algorithm Issue

A group of results for one period (roughly July to October 1984) appears to indicate a reporting 

problem with the dosimetry algorithm used to calculate dose equivalents. In general, these results 

contain a gamma component that was calculated to be zero and a neutron dose that was calculated 

to be between about 15 and 50 mrem. However, the deep dose on both the report cards and in the 

electronic record was zero.

A review of a paper copy of the dose algorithm from that time (Author unknown undated) and 

discussion with the algorithm developer indicated that the algorithm was developed in such a way that 

it should not have been possible to have a zero gamma dose with a nonzero neutron dose. In such a 

case, the algorithm would set the neutron dose to zero.

In these cases, however, the deep dose is reported as zero, and the neutron component was not set 

to zero before it was reported.

6.3.2.2.2 Possible Manual Correction

In another group of records, the deep dose is much greater than the sum of the gamma and neutron 

components. In the electronic data, these records appear during a period identified as 1976. A 

review of a number of these records found, in all cases, a letter in the file instructing the staff to modify 

the individual’s data due to a dose reconstruction or reevaluation. It appears that dose components 

were not provided in the letter and, therefore, were not made to add up to the deep dose.

The 1976 date in the electronic record appears to have no relationship to the actual date associated 

with the dose record. According to the reports, many of the actual doses were assigned from 1984 to 

1986.

6.3.2.3 Shallow Dose Minus Neutron Dose is Less than Photon Dose

The shallow dose is reported in the Dosimetry History by Individual. After 1976, if the reported 

shallow dose, after subtracting the neutron dose, is less than the photon dose, the shallow dose 

should be assumed to be zero. 

6.3.3 Database Table-Specific Issues

Two database tables contain the external dosimetry data in RHRS, as discussed in the following 

sections. Each table has specific information on the external monitoring period, and the distinctions 

between the tables are notable.

6.3.3.1 RHRST_ED_TLD_HISTORY

This table contains external dosimetry data for years generally before 1991, the time of RHRS 

implementation. These data migrated from earlier computer records systems such as the HSDB. 

Most of the records contain only a date referred to as “Activity Date.” In general, this Activity Date is 

close to the dosimeter return date if the actual return date is available.
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To migrate these data to the HIS-20 electronic database, an issue date had to be fabricated. 

Because the Activity Date is closer to the return date and there was no information on the exchange 

frequency, the issue date was set to 1 d before the return date.

6.3.3.1.1 1976 Records (Individual Employed After 1976)

This table contains a record dated December 31, 1976, for every individual in the database who was 

hired before 1989, even if they did not start work until after that date. This appears to have been an 

artifact from the initial migration of data from HSDB to RHRS. Therefore, a data record for 1976 might 

appear in Health Physics file reports called External Dosimetry (TLD) Detail (from RHRS) and 

Dosimetry History by Individual (from HIS-20) when the individual was not yet hired.

Zero Dose Records

As a general rule, these records are not attributed to the individual, and they report a deep dose of 

zero.

Nonzero Dose Records

A 1976 record appears occasionally with a deep dose greater than zero. Such records are regarded 

as valid, and the official dose is attributed to the individual even though it is outside the employment 

period (see Section 6.3.2.2.2).

6.3.3.1.2 1976 Records (Individual Employed Before 1976)

For individuals employed before 1976, the 1976 record represents a lump sum total of the deep dose 

for all previous years. However, the details for each year should be available during a review of report 

cards for an individual.

In addition, a database from the Colorado Department of Health was used to replace the lump sum 

with an annual deep dose value (Ruttenber et al. 2003). Again, there is no electronic source for the 

deep dose components (neutron and gamma) or for skin and extremity values.

6.3.3.1.3 Post-1976 Records

Because the only date available before 1991 was the Activity Date, records can appear in reports that 

are outside the employment period. The Activity Date was used to document a “wear period” if there 

was no knowledge of the frequency of the dosimetry exchange. Therefore, the records might appear 

before the hire date or after the termination date.

6.3.3.2 RHRST_ED_TLD_DOS

This table, which has a structure identical to RHRST_ED_TLD_HISTORY, contains post-1991 data. 

The records result from a download of the external dosimetry computer system called FALCON. This 

system collects and processes data directly from the Panasonic TLD readers. The records generally 

contain values for each column, including a variety of dates such as issue date, return date, and 

activity date.

There could be discrepancies between the monitoring period and the employment period. Individuals 

who did not check out properly might not have an accurate employment termination date. In addition, 

the computer systems typically documented the dates that the person wore dosimetry rather than the 

employment period. This is particularly true for subcontractors.
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6.3.3.2.1 Dose History Hard-Copy File Contents

The RFP Radiological Health organization reviewed the individual dose record and summarizes it in 

an Occupational Dose Report worksheet (Attachment A, Figures A-1 and A-2). This document shows 

the measured dose on an annual basis and summarizes the available dose data from the printed 

record in the rest of the file. These data were compared with the computerized data, which are in the 

Dosimetry History by Individual report (Figures A-3, A-4, and A-5). Before 1976, the data were 

entered on an annual basis. After 1975, this report provided a dosimeter-by-dosimeter reading. The 

End Date indicates the end of the wear period. Comparison with the previous End Date can indicate 

the exchange frequency. If the Begin Date was not known, it was set to 1 d before the End Date. In 

this case, it can be assumed that the badge was worn from approximately the day after the previous 

End Date to the indicated End Date for that period. 

Several other reports are included, some of which contain more dosimetry result detail. The following 

observations are from Savitz (undated) and the result of review of these records by James M. 

Langsted (ORAUT 2018a): 

• Early years are reported on the Health Physics External Exposure Run, which provides a 

quarterly breakdown. Even though dosimeters might have been exchanged more frequently, 

data are summarized by quarter and more detailed data are not included.

• The 1953 to 1958 report Health Physics External Exposure Activity Run Yearly (Figure A-6) 

contains a quarterly summary of the exposure data for an individual. The dose equivalent 

values reported are Skin, Pen (penetrating; the deep dose to the whole body), and Hand 

(regarded as the dose to the extremity, if monitored).

• The 1959 to 1963 report Health Physics Yearly External Exposure Run (Figure A-7) contains 

all details for each measurement for an individual. Each reading is on a separate line, which 

reveals the frequency of the monitoring. The dose equivalents are reported as Skin, Penet 

(the deep dose to the whole body), and Wrist (the dose to the extremity, if monitored).

• The 1964 report Health Physics External Exposure Activity Run Yearly (Figure A-8) appears to 

be a transition report. It contains a quarterly summary of exposure data for an individual. The 

dose equivalent values reported are Skin, Pen (the deep dose to the whole body), and Hand 

(the dose to the extremity, if monitored). 

• The External Dosimetry (TLD) Detail, Computerized Information Through xx-xx-xx or External 

Dosimetry (TLD) Detail, Computerized Information for CY (calendar year) 19xx report 

(Figure A-9) provides dosimeter reading detail for the years indicated. The Activity Date 

indicates the nominal (a few days to either side) end date of the dosimeter wear period. In the 

context of this report, Time Code indicates the identified exchange period for the badge:

– Time Code 1, semimonthly (twice per month)

– Time Code 2, monthly

– Time Code 3, annual

– Time Code 4, quarterly

For the period this report was used, the shortest routine exchange period was semimonthly as 

indicated and not biweekly as discussed in Section 6.4.2.

Time Codes are typically seen in dosimetry records from the late 1970s through the late 

1980s. For 1990, where there is a single badge with an end of the year date, or the end of 

employment date for the year, the DR should assume an annual badge exchange. Note that 
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for 1990, there is typically no Time Code listed for the dosimeter result. After review of claim 

data, it was determined that time codes after 1990 should not be used for determining a badge 

exchange frequency because they are no longer accurately entered. 

During the transition between the Harshaw and the Panasonic badges, RFP used a code to 

indicate the source of the dosimetry result (ORAUT 2018a): 

– Type code C (calculated): Panasonic badge result (calculated in Panasonic computer 

system), no wrist dosimeter data

– Type code R (raw): Harshaw badge chip readings (raw chip readings, result calculated 

in RHRS database system), no wrist dosimeter data

– Type code H (hybrid): Panasonic badge result and Harshaw wrist dosimeter chip 

readings

• The Health Physics External Radiation Exposure Report for Year XX (also known as report 

card) (Figure A-10) provides quarterly totals for the year. Because dose limits were on a perquarter basis, the purpose of this report was to monitor compliance with these limits. The 

dosimeter detail was lost. 

• The 1965 to 1989 Health Physics External Radiation Exposure Report contains a quarterly 

summary of exposure data for an individual. The dose equivalent values reported are Pen 

(the deep dose to the whole body), Skin, and Hand (the dose to the extremity, if monitored). In 

addition, these reports contain a lifetime (career) deep dose for exposure at RFP. After 1976, 

a column was added to the report for a value described as Forearm. This dose equivalent 

appears to be similar to that for the hand. In 1977, the dose to the hand was set to the greater 

of the skin of the whole body and the measurement calculated from the actual wrist dosimeter 

(Falk 1976). 

• For individuals employed after 1976 and until about 1986, there might occasionally be a report 

called External Dosimetry (TLD) Detail. This report contains greater detail on each 

measurement made during this period and a breakdown of gamma and neutron components.

• The Radiation Dosimetry Individual Lifetime Report (Figure A-11) provides very little detail 

other than a verification of the Reported Lifetime Dose. This includes offsite doses (from 

previous employers), which should be detailed in the file. 

• The RHRS Data report (Figure A-12) provides details of the dosimeter results. The advantage 

of this report is that it shows the breakdown of the deep dose into neutron and gamma 

components.

• The Radiation Dosimetry Detail Report, Termination Report (Figures A-13 and A-14) provides 

a verification of lifetime and post-1987 exposure.

• The Occupational Radiation Exposure Information (Figure A-15) provides annual Whole Body, 

Hand, Forearm, and accumulated RFP whole-body (ACCUM AT RF) doses. The whole-body 

dose is assumed to be penetrating.

These data enable compilation of an external dosimetry history, as follows:

• 1951–1976. Quarterly dose history (RHRS data will provide a neutron/gamma breakdown), 

• 1959–1964. Dosimeter exchange history, and
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• 1977–2005. Dosimeter exchange history. 

In some cases, additional data are available. The dose reconstructor is responsible for using the 

information in this TBD to provide assumptions favorable to claimants to fill in unavailable detail for a 

claimant's external dosimetry record.

6.3.4 Interpretation of Dosimetry Data

Table 6-1 provides detail for the interpretation of the values, zeros, and blanks encountered in the 

RFP reports detailed in the previous section.

6.3.5 Additional Data Available

There are additional sources of information, which are known to exist (ORAUT 2018a), that contain 

detail that is not in the dose history file. These data might provide detail useful to refining dose 

estimates for some workers.

6.3.5.1 Rocky Flats Work History File

The RFP Human Resources department kept job assignment records for many years on 5- by 7- in. 

cards (ORAUT 2018a). Images of these cards could provide a further indication of the type(s) of work 

performed by the worker. This information is not in the dose history file.

6.3.5.2 Neutron Dose Reconstruction Project File

The NDRP provided an updated assessment of the neutron exposures that monitored workers 

received while performing work in the RFP plutonium production facilities from 1952 to 1970. The 

NDRP reassessed the neutron doses either by rereading neutron films and plates used to monitor 

workers for neutron exposures or by estimating the neutron doses for periods when a worker was not 

monitored for neutron exposures while working in a plutonium-related building. The focus of the 

NDRP was neutron dose, so the study contains data primarily on plutonium workers and not on 

uranium and other workers who were unlikely to be monitored for neutron exposure. The study has 

provided NDRP-generated results for those workers in the study for whom there are EEOICPA claims. 

These data are described in the NDRP protocol document (Falk et al. 2005).

When records indicate an energy employee worked in a plutonium facility before 1971, and no NDRP 

data exists in the dosimetry records, additional effort should be made to verify the existence or 

absence of NDRP data for a specific energy employee.

6.3.5.3 Job Exposure Matrix

A DOE-funded study performed by the University of Colorado Health Sciences Center and the 

Colorado Department of Public Health and Environment (Ruttenber et al. 2003) developed a Job 

Exposure Matrix that identified the building assignment and a job title snapshot during September for 

each year from 1952 to 1989. This matrix was matched with external dosimetry results, and it could 

provide dose distributions for groups and job titles to assist in estimating dose for unmonitored 

workers. On April 4, 2006, NIOSH reviewed the data available from this project and concluded that 

the material is valuable for epidemiological studies but is of limited utility for NIOSH dose 

reconstruction.
 

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Table 6-1. Interpretation of reported data (ORAUT 2018a).

a

Report Reported quantity Interpretation of zeros

Interpretation of 

blanks (no data)

Individual and annual 

data

Monitored/

unmonitored

Occupational Dose 

Report

Hand-generated 

summary of dosimetry 

record (see Figures A-1 

and A-2 and Section 

6.3.3.2)

Annual totals in mrem 

according to Tables 6-

10 and 6-17 and 

Section 6.8.2.1.

Deep dose

Extremity dose

Skin dose (see Section 

6.3 and 6.6.1.2).

Zero indicates a 

monitored exposure 

reported as zero.

Blank indicates 

unmonitored during 

that period. Form does 

not indicate if individual 

was onsite during 

unmonitored period.

See Section 6.3.2. Evidence is that at 

RFP, if employee was 

monitored, results were 

reported. Between 

1964 and the early 

1990s, all onsite 

individuals were 

monitored with a body 

badge. See Section 

6.4.1.

Dosimetry History by 

Individual

Computer-generated 

summary of dosimetry 

(see Figures A-3, A-4, 

and A-5 and Section 

6.3.3.2) See Note a. 

below

Annual total through 

1976 and individual 

dosimeter results 

thereafter in mrem (see 

references above)

DDE

SDE-SK (skin)

SDE-EX (extremity)

Neutron (neutron is

included in DDE and 

SDE before 1977).

LDE–irrelevant to dose 

reconstruction.

Before 1977, a zero in 

the neutron field should 

be disregarded. During 

this period, the neutron 

dose is included in both 

the DDE and the SDE 

and is not available 

separately. After 1976, 

the neutron dose is 

reported in the Neut. 

column and is included 

in the SDE values. 

Otherwise, a zero 

indicates a monitored 

exposure reported as 

zero.

Blank indicates a 

nonreported value for 

that period or 

dosimeter exchange.

See Section 6.3.2. Entries are not 

provided for periods 

when the individual 

was not employed at 

RFP. A missing result 

in a series of 

continuous dosimetry 

results is probably the 

result of a missed 

dosimeter exchange. 

See Section 6.5.3.
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Report Reported quantity Interpretation of zeros

Interpretation of 

blanks (no data)

Individual and annual 

data

Monitored/

unmonitored

Health Physics 

External Exposure 

Activity Run Yearly 

(Figure A-6)

Quarterly total of dose 

in mrem according to 

Tables 6-10 and 6-17 

and Section 6.8.2.1

Penetrating dose, 

Extremity dose (see 

Section 6.10),

Skin dose (see 

Sections 6.3 and 

6.6.1.2). These totals 

might result from 

multiple dosimeter 

exchanges during the 

quarter.

Zero indicates 

monitored dose 

reported as zero. 

Blank indicates 

dosimetry result was 

not measured for that 

period. Extremity 

dosimeters were not 

worn by all individuals. 

See Section 6.3.2.

Annual Hand totals are 

based only on 

measured Hand values 

for that year.

Blanks indicate that 

individual was not 

monitored for that 

dose.

Health Physics Yearly 

External Exposure 

Run

Dosimeters exchanged 

biweekly (Figure A-7)

(see codes indicated 

below)

Units are mrem 

according to Tables 6-

10 and 6-17 and 

Section 6.8.2.1.

Zero or blank indicates 

no measured dose 

under that filter

For gamma/beta and 

gamma/X-ray badges 

(Codes 1 and 2), a 

blank result indicates a 

zero reading. This is 

based on a review of 

records of this type 

showing only positive 

and blank readings for 

a badge. For neutron 

badges (Code 3), 

neutron dose is placed 

in B/CD column and 

other columns are left 

blank because they are 

not used.

Sheet shows only 

individual dosimeter 

results.

“Type 0” in fifth column 

seems to indicate that 

neutron dosimeter was 

either lost, unreadable, 

or below the detection 

limit. This has not 

been determined. 

Assumptions favorable 

to claimants should be 

made.

Health Physics 

External Exposure 

Activity Run, Yearly

(Figure A-8 and 

Section 6.3.3.2)

Quarterly totals are 

mrem according to 

Tables 6-10 and 6-17 

and Section 6.8.2.1.

Zeros indicate total for 

dosimeters all reporting 

zero

Blank indicates 

individual was not 

monitored during that 

quarter

This is a summary 

report. If individual 

results are available, 

they should match.

A blank indicates that 

the individual was not 

monitored during that 

period, either because 

the worker was not 

onsite or was not 

expected to exceed 

some currently 

applicable 

administrative limit.
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Report Reported quantity Interpretation of zeros

Interpretation of 

blanks (no data)

Individual and annual 

data

Monitored/

unmonitored

External Dosimetry 

(TLD) Detail

(Figure A-9 and 

Section 6.3.3.2)

Individual dosimeter 

results are mrem 

according to Tables 6-

10 and 6-17 and 

Section 6.8.2.1. 

Dosimetry results 

calculated as indicated 

in Section 6.3.

Doses reported down 

to zero. Zero indicates 

dosimeter response 

less than background 

value used.

Blank indicates an 

unusual situation where 

part of dosimetry result 

is not available. If 

dosimeter result is not 

available, no entry will 

be recorded.

Individual dosimeter 

results are totaled for 

quarter or CY on other 

reports.

Between 1964 and the 

early 1990s, all onsite 

individuals were 

monitored with a body 

badge. See Section 

6.4.1. If dosimeter 

result is missing, either 

individual did not 

exchange badge or 

was not a site 

employee.

Health Physics 

Annual External 

Radiation Exposure 

Report for Year XX

(Figure A-10)

Quarterly totals are 

mrem according to 

Tables 6-10 and 6-17 

and Section 6.8.2.1. 

Dosimetry results 

calculated as indicated 

in Section 6.3.

For body badge results, 

zeros indicate a sum of 

zeros reported for all 

external dosimetry 

results during that 

quarter. It is likely that 

individuals were 

monitored with a body 

badge but did not 

receive an extremity 

dosimeter. In this 

case, a zero in the 

Hand column indicates 

the individual was likely 

unmonitored for 

extremity dose.

A blank indicates that a 

dosimetry result was 

not obtained for that 

period. This could be 

because the individual 

was not employed, not 

monitored, or did not 

submit a badge during 

that period. A blank in 

the Hand column 

indicates that the 

individual was not 

monitored for extremity 

dose.

This is a rollup of 

dosimetry results for 

quarter and should be 

consistent with annual 

dose reported 

elsewhere.

If a zero is reported, it 

is a result of external 

dosimetry results of 

zero. After 1976, if the 

hand dose equals the 

skin dose, this 

indicates that the hand 

was not separately 

monitored and the skin 

dose was used to 

estimate the hand 

dose. Before 1977, the 

measured hand dose 

was reported.

Radiation Dosimetry 

Detail Report, 

Individual Lifetime 

Report

(Figure A-11)

This report shows only 

deep (Pen) dose in 

mrem according to 

Tables 6-10 and 6-17.

A zero indicates that 

external dosimetry 

measurements were 

performed resulting in a 

total of zero.

Blanks indicate no 

external dosimetry 

measurements were 

performed. 

Occupational exposure 

from other facilities is 

available only if 

reported to RFP. Often 

other facility exposure 

records were not 

available.

This is a rollup of data 

for the period indicated. 

It includes other facility 

exposure if available. 

Observed data 

discrepancies as 

indicated in Section 

6.3.2 are possible in 

these totals. 

For unmonitored 

individuals, fields will 

show a blank when no 

external dosimetry 

measurements were 

recorded.
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Report Reported quantity Interpretation of zeros

Interpretation of 

blanks (no data)

Individual and annual 

data

Monitored/

unmonitored

Radiation Health 

Records System –

View TLD Data

(Figure A-12)

Individual dosimeter 

results are in mrem 

according to Tables 6-

10 and 6-17 and 

Section 6.8.2.1. 

Dosimetry results 

calculated as indicated 

in Section 6.3.

Time Code and Type 

are as explained in 

Section 6.3.3.2.

Zeros in all fields 

except background 

(BK-1 and BK-2) 

indicate a measured 

dosimetry result of 

zero. Zeros in the 

background fields are 

irrelevant for Type C 

records and, for Type H 

or Type R records, 

indicate the 

background values that 

have been used in 

correcting the reported 

dosimetry results.

It is not clear if blanks 

are present on this 

report. If they do exist, 

they should be 

interpreted as no 

external dosimetry 

measurement was 

recorded for that 

period.

These individual results 

are rolled up into 

annual totals.

The Activity Date 

indicates the 

approximate end of the 

dosimeter wear period. 

This date, used with 

the Time Code 

(exchange frequency) 

indicates the presence 

of missing dosimeters. 

A gap indicates a lost 

dosimeter, a dosimeter 

worn for multiple 

periods, or a period for 

which the individual 

was not monitored.

Radiation Dosimetry 

Detail Report, 

Termination Report

(Figures A-13 and 

A-14)

This report shows only 

deep (Pen) dose in 

mrem according to 

Tables 6-10 and 6-17.

Zeros in the data for a 

specific year indicate 

an external dosimetry 

measurement of zero.

If the internal and 

external data for a 

specific year are blank, 

the zero in the TEDE 

and TODE fields for 

that year are incorrect 

and should be blank.

A zero in the 

Cumulative or Lifetime 

fields indicates that all 

measurements 

contributing to that total 

are zero. 

In the data for specific 

years, blanks indicate 

that no external 

dosimetry 

measurements were 

recorded for that year. 

This is a rollup of 

dosimetry results that 

might be available 

elsewhere in worker 

external dosimetry 

record files. Observed 

data discrepancies as 

indicated in Section 

6.3.2 are possible in 

these totals.

For completely 

unmonitored 

individuals, the 

Cumulative and 

Lifetime external dose 

fields show zeros. 

Unmonitored periods 

would be undetectable 

in this report.
 

Document No. ORAUT-TKBS-0011-6 Revision No. 0

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9 Page 24 of 109

Report Reported quantity Interpretation of zeros

Interpretation of 

blanks (no data)

Individual and annual 

data

Monitored/

unmonitored

Occupational 

Radiation Exposure 

Information

(Figure A-15)

This report shows dose 

in rem according to 

Tables 6-10 and 6-17. 

Penetrating dose is 

reported for whole body 

and extremity dose is 

reported as both 

forearm and hand. 

Extremity dosimetry is 

further explained in 

Section 6.8.2.1.

Zeros indicate an 

annual total of zero 

based on both external 

dosimetry results of 

zero and unmonitored 

periods. 

Blanks should not be 

present on this report.

This is a rollup of 

dosimetry results that 

might be available 

elsewhere in worker 

external dosimetry 

record files. Observed 

data discrepancies as 

indicated in Section 

6.3.2 are possible in 

these totals.

This report includes 

both monitored and 

unmonitored periods. It 

is impossible to 

determine the 

unmonitored periods 

from this report.

a. LDE = lens of the eye dose equivalent; SDE = shallow dose equivalent; TEDE = total effective dose equivalent; TODE = total organ dose equivalent.

Report codes for Health Physics Yearly External Exposure Run

Code Explanation

1 Gamma/beta dosimeter

2 Gamma/X-ray dosimeter

3 Neutron dosimeter

Dose column labels

Code Explanation

B/CD Body/cadmium

B/BR Body/brass

B/OW Body/open window

W/BR Wrist/brass

W/OW Wrist/open window

SKIN Skin dose

PENET or PEN Penetrating dose (equivalent to deep dose)

WRIST Extremity dose (as measured by wrist 

dosimeter)
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6.4 HISTORICAL ADMINISTRATIVE PRACTICES

6.4.1 Badged Population

When plant operations began in 1951, there was no external dosimetry, and there was not much 

radioactive material at the Plant. In September 1952, dosimeters became available for use. Some 

individuals in Building 9911 received neutron dosimeters. The use of dosimetry expanded to other 

RFP production operations. 

For some radiation workers, no neutron monitoring at all was performed during the period from 1952 

through 1970. For other workers, from 1967 to 1970, nuclear track emulsion, type A (NTA) film 

badges were issued but not evaluated after they were used (Falk et al. 2005). 

Two analyses were performed to indicate the portion of the plant population that was monitored using 

external dosimetry. The results are shown in Figure 6-7. The solid line indicates a manual analysis 

that was performed on the data in all of the NIOSH EEOICPA claim files that were available in 

October 2005. At that time, 1,046 claimant files were available for analysis. The broken line 

indicates a computer analysis that was performed on the RFP external dosimetry database. Over 

288,000 employee-years of data were evaluated. This shows the portion of the plant population that 

was monitored. 

Figure 6-7. Portion of plant population badged. 

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

1951 1961 1971 1981 1991 2001

Year

Badged

EEOICPA Claims

Plant Population

A steady increase occurred until 1964, when the security badge was incorporated in the dosimetry 

badge, which ensured that each individual wore a dosimetry badge (Putzier 1982). This design was 

maintained until 1991 (Jens 1990), when the security badge was separated from the dosimeter and 

individuals unlikely to receive occupational radiation exposure greater than 100 mrem/yr were no 

longer issued dosimeters. The dip in 1969 is probably a result of the personnel displacement from the 

Building 776 fire. The disparity between the EEOICPA claim data and the population data is thought 

to be the result of multiple hires and terminations that were not accurately recorded in the electronic 

 

1 In the early years, two-digit building numbers were used. These were later changed to corresponding three-digit numbers. 

For example, Building 81 became Building 881. Buildings 371 and 771 present the only case in which there could be 

confusion, but the change to three-digit numbers took place well before Building 371 was built; therefore, Building 71 

always refers to Building 771.
Document No. ORAUT-TKBS-0011-6 Revision No. 03 Effective Date: 01/14/2019 Page 26 of 109

 

data. When the claim files are reviewed, these data are refined based on the paper records and a 

more accurate data set is used for the analysis. The reduction in badging that began in 1991 is the 

result of an effort by the site Radiological Protection organization to identify personnel unlikely to 

exceed the exposure criteria for radiological workers (100 mrem/yr) and to discontinue badging of 

those personnel. The increase in 1998 was the result of rebadging personnel to perform 

decontamination and decommissioning (D&D) work, and the reduction at the end was the result of 

discontinuing badging after the completion of D&D work (which could result in significant worker 

dose).

For some plutonium workers, neutron monitoring was not provided until the early 1960s, and their 

doses of record might not include significant contributions from neutron exposure received before 

being issued a neutron dosimeter. These workers included most of the employees working in 

Building 71 (now Building 771). Only a small number (10 to 18) of these employees were monitored 

for neutron exposure, and that monitoring occurred only from October 1956 to September 1957 (Falk 

et al. 2005).

A group of plutonium workers [the plutonium metal (foundry) workers in Building 71] was not 

monitored for whole-body, penetrating gamma, and X-ray doses until February 1957. Instead, they 

were issued only a wrist dosimeter (Falk et al. 2005). 

The average dose trend for monitored individuals is shown in Figure 6-8. This trend is influenced by 

the number of workers monitored. For example, when only some workers were monitored, those 

selected generally had the highest potential for exposure, and the average would be higher. When all 

site employees were monitored, the average was reduced by those monitored employees who did not 

work in the production areas.

Figure 6-8. Average measured worker dose (data from ORAUT 2007a).

0.000

0.500

1.000

1.500

2.000

2.500

3.000

1950 1955 1960 1965 1970 1975 1980 1985 1990 1995 2000 2005

Year

Average Worker Dose (rem)

Pen

Skin

6.4.2 Badge Exchange Frequency

The determination of badge exchange frequencies was based on the potential for external dose and 

the necessity to control dose to administrative limits. Badges were exchanged at various frequencies. 

Early dosimetry was exchanged on a weekly basis, which later became biweekly (as illustrated in 

Figure A-7), semimonthly (twice per month), and monthly. It is not clear when the change from 

biweekly to semimonthly occurred. Before 1991, dosimetry was generally exchanged on 
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semimonthly, monthly, quarterly, and annual frequencies. After 1990, exchange frequencies 

generally went to monthly, quarterly, and semiannually.

Badge exchange frequency records have not been maintained. If individual dosimeter readings were 

maintained, the exchange frequency for an individual can be determined by reviewing the external 

dose record. After 1976, the dose record shows a dosimeter reading for each exchange. For earlier 

years, the dose has been combined into quarterly records for which the exchange frequency has been 

lost, although it is reasonable to assume that badges were exchanged at least quarterly (see 

Figures A-3 to A-15 as documented in Section 6.3.3.2).

To determine the exchange frequencies used before 1976, original dosimetry laboratory worksheets 

were reviewed (ORAUT 2018a). Many of these worksheets have been assembled as part of the 

NDRP. Dosimetry laboratory worksheets from 1951 to 1970 were assembled and organized. A 

sample was obtained during preparation of this report by selecting the September folder for each 

year. A review of each worksheet determined the exchange frequency, building, and dosimeter type 

(photon, beta, or neutron). These data were organized and reviewed to determine the most frequent 

exchange for the major job categories (see Attachment B) by year. The worksheets do not indicate 

job assignment. It was necessary to evaluate the job category based on the building and exchange 

frequency. In cases where multiple exchange frequencies were indicated for a major job category, 

the more frequent exchange frequency was selected. This provides an assumption favorable to 

claimants when determining missed dose. Dosimetry worksheets are not readily available for 1970 to 

1976, so exchange frequencies were extrapolated forward for those years. Table 6-2 lists the results 

of this analysis. These are the default values to use if the exchange frequency cannot be determined. 

If no job category can be determined, the dose reconstructor should use the most frequent exchange 

rate for that year. For semimonthly badge exchanges, biweekly exchange should be assumed 

(26 exchanges per year instead of 24) when an approach favorable to claimants is desired.

In 1991 and 1992, the Dosimetry History by Individual may have actual badge begin and end dates for 

the dosimeters during that time. In some cases, the Dosimetry History by Individual will report 

sequentially numbered begin and end dates (i.e., 30-MAR-1991 and 31-MAR-1991). There could also 

be other dosimetry records indicating an “Activ_Date” or “Activity_Date” with, typically, March, June, 

September, and December badges. In these cases, when actual badge begin and end dates are not 

provided, a quarterly exchange frequency should be assumed.

6.4.3 Field-Specific Calibration Factors

Film dosimeters required the use of workplace-specific calibration factors, so it was necessary to 

know the facility in which the individual worked (ORAUT 2018a). Individuals sometimes worked in 

other facilities on temporary or overtime assignments, which the Dosimetry department could not 

detect. Area-specific calibration factors were necessary to evaluate readings from the X-ray/gamma 

dosimeters used in the plutonium areas and the beta/gamma dosimeters used in the uranium areas. 

Exposure of the dosimeter in a different field could not be detected, which introduced a source of 

uncertainty.

TLD systems use more tissue-equivalent detection elements (ORAUT 2006a, Section A.2.1.2), which 

do not require a field-specific calibration factor. This source of uncertainty is minimal with these 

dosimeters. 

6.4.4 Minimum Reported Dose

RFP appears to have embraced a philosophy of reporting dose down to zero between 1951 and 

1992 (ORAUT 2018a). In 1993, the Plant adopted a minimum reported dose threshold to remove the 

bias associated with reporting low doses and truncating doses calculated to be small negative 
Document No. ORAUT-TKBS-0011-6 Revision No. 03 Effective Date: 01/14/2019 Page 28 of 109

 

numbers to zero. In 1993, a minimum reported dose level of 10 mrem was adopted. Any dose below 

this level was reported as zero (KHC 2001). This policy is consistent with the limits of detection 

(LODs) reported elsewhere in this TBD.

6.4.5 Recorded Dose Practices

Table 6-3 provides a summary of the calculations used to determine the recorded dose at RFP.

6.5 COMMON ISSUES

This section discusses issues common to external photon, neutron, and electron dose measurement 

at RFP. These issues are addressed further only if there is an issue specific to that type of dose 

measurement.

6.5.1 Number of Zero Readings

At present, available dosimetry records do not consistently provide individual dosimeter results for all 

of the early years. Therefore, it is often necessary to estimate the dosimeter exchange frequency for 

some or all of the period from 1951 to 1976. Table 6-2 provides an estimate based on major job 

category. If an individual’s job assignment cannot be determined, the most frequent dosimeter 

exchange used during that year should be assumed. This assumption is favorable to claimants.
 

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Table 6-2. Conservatively determined default dosimeter exchange frequencies.

a,b

Year

Chemical

operators, 

Pu

Chemical

operators,

U

Metallurgical

operators, 

Pu

Metallurgical 

operators,

U

Maintenance 

workers

Support 

personnel

Analytical 

laboratory 

technicians

Site 

support 

personnel

Radiation 

control 

technicians

D&D 

workers

1951 Biweekly Biweekly Biweekly Biweekly Monthly Monthly Biweekly Monthly Biweekly Monthly

1952 Biweekly Biweekly Biweekly Biweekly Monthly Monthly Biweekly Monthly Biweekly Monthly

1953 Biweekly Biweekly Biweekly Biweekly Monthly Monthly Biweekly Monthly Biweekly Monthly

1954 Biweekly Biweekly Biweekly Weekly Monthly Biweekly Biweekly Monthly Biweekly Biweekly

1955 Biweekly Biweekly Biweekly Weekly Monthly Biweekly Biweekly Monthly Biweekly Biweekly

1956 Biweekly Biweekly Biweekly Weekly Monthly Biweekly Biweekly Monthly Biweekly Biweekly

1957 Biweekly Biweekly Biweekly Weekly Monthly Biweekly Biweekly Monthly Biweekly Biweekly

1958 Weekly Weekly Weekly Weekly Monthly Biweekly Biweekly Monthly Weekly Biweekly

1959 Weekly Biweekly Weekly Weekly Monthly Biweekly Biweekly Monthly Weekly Biweekly

1960 Weekly Weekly Weekly Weekly Monthly Monthly Biweekly Monthly Weekly Monthly

1961 Biweekly Biweekly Biweekly Weekly Monthly Monthly Weekly Monthly Biweekly Monthly

1962 Biweekly Biweekly Biweekly Weekly Monthly Monthly Weekly Monthly Biweekly Monthly

1963 Biweekly Monthly Biweekly Biweekly Monthly Monthly Quarterly Monthly Biweekly Monthly

1964 Biweekly Monthly Biweekly Monthly Quarterly Monthly Quarterly Quarterly Biweekly Monthly

1965 Monthly Not Applicable Monthly Monthly Quarterly Quarterly Quarterly Quarterly Monthly Quarterly

1966 Monthly Not Applicable Monthly Monthly Monthly Quarterly Quarterly Quarterly Monthly Quarterly

1967 Biweekly Not Applicable Biweekly Monthly Monthly Quarterly Quarterly Quarterly Biweekly Quarterly

1968 Biweekly Not Applicable Biweekly Monthly Monthly Monthly Monthly Monthly Biweekly Monthly

1969 Biweekly Not Applicable Monthly Monthly Monthly Monthly Monthly Monthly Monthly Monthly

1970 Biweekly Not Applicable Monthly Monthly Monthly Monthly Monthly Monthly Monthly Monthly

1971 Biweekly Not Applicable Monthly Monthly Monthly Monthly Monthly Monthly Monthly Monthly

1972 Biweekly Not Applicable Monthly Monthly Monthly Monthly Monthly Monthly Monthly Monthly

1973 Biweekly Not Applicable Monthly Monthly Monthly Monthly Monthly Monthly Monthly Monthly

1974 Biweekly Not Applicable Monthly Monthly Monthly Monthly Monthly Monthly Monthly Monthly

1975 Biweekly Not Applicable Monthly Monthly Monthly Monthly Monthly Monthly Monthly Monthly

1976 Biweekly Not Applicable Monthly Monthly Monthly Monthly Monthly Monthly Monthly Monthly

a Biweekly (every 2 weeks), assumed because it is favorable to claimants over semimonthly (twice per month).

b. Source: Study described in Section 6.4.2.
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Table 6-3. Summary of historical recorded dose practices.

a

Period and dosimeter Dosimeter measured quantities Compliance dose quantities

1951–1956, 

Two-element film (photon) 

+ track plate (neutron)

Owdose = (Owdensity – Cddensity) × CFnet Ow

Cddose = Cddensity × CFCd

Ndose = neutron tracks × CFneutron

Pen = Cddose + 0.5 × Owdose + Ndose

Skin = Cddose + Owdose + Ndose

1957–1959, 

Two-element film + NTA 

film

Owdose = (Owdensity – Cddensity) × CFnet Ow

Cddose = Cddensity × CFCd

Ndose = neutron tracks × CFneutron

Pen = Cddose + 0.5 × Owdose + Ndose

Skin = Cddose + Owdose + Ndose

5/1953–10/1970, 

Two-element film (beta)

Owdose = (Owdensity – Cddensity) × CFnet Ow

Cddose = Cddensity × CFCd

(no neutron measured)

Pen = Cddose

Skin = Cddose + Owdose

3/1960– (Bldg. 71), 

1/1963– (other Pu), 2/1968 

(Bldg. 81 and 91)

 –1962, 

Three-element film 

+ NTA film

Owdose = (Owdensity – f × Brdensity) × CFnet Ow

Brdose = (Brdensity – Cddensity) × CFnet Br

Cddose = Cddensity × CFCd

Ndose = neutron tracks × CFneutron

Pen = Cddose + Brdose + 0.35 × Owdose + 

Ndose

Skin = Cddose + Brdose + Owdose + Ndose

1963–1969, 

Multiple-element film 

+ NTA film

Owdose = (Owdensity - f × Brdensity) × CFnet Ow

Brdose = (Brdensity - Cddensity) × CFnet Br

Cddose = Cddensity × CFCd

Ndose = neutron tracks × CFneutron

Pen = Cddose + Brdose + 0.35 × Owdose + 

Ndose

Skin = Cddose + Brdose + Owdose + Ndose

1969–1970

1/1970, (Bldg. 771), 

4/1970 (other Pu Bldgs) 

 –10/1970 (all others)b 

TLD + NTA film

Gdose = PTLD

Sdose = STLD

Ndose = neutron tracks × CFneutron

Pen = Gdose + Ndose

Skin = Sdose + Ndose

1971–1982

TLD-700 + TLD-600/700

Gdose = PTLDc

Sdose = STLD

Ndose = determined from albedo algorithm

Pen = Gdose + Ndose

Skin = Sdose + Ndose

if Skin < Pen, then Skin = Pen

1983–1989, Panasonic 

UD-802 + UD-809

Photon deep

Photon/Beta shallow

Neutron

Pen = photon deep + neutron

Skin = photon/beta shallow + neutron

1990–2004, Panasonic 

UD-802 + UD-809

Hs,gamma 

Hd,gamma

Hs,beta

Ndose 

Hd = Hd,gamma + neutron

Hs = Hs,gamma + Hs,beta + neutron

a. Owdensity = OW (measured density); Owdose = OW (determined dose); Cddensity = cadmium filter (measured density); 

Cddose = cadmium filter (determined dose); Brdensity = brass filter (measured density); Brdose = brass filter (determined 

dose); f = factor to correct for brass attenuation of X-rays (1.14 or 1.17 or nonlinear factor); Ndose = neutron dose; CF = 

calibration factor determined from calibration films.

b. Except some groups in Building 444 and miscellaneous other groups.

c. Average of two crystals, or one crystal if one crystal is zero.

Once the estimated exchange frequency has been established, the number of zero readings must be 

estimated. For the majority of the time, estimates of zero readings can be obtained using actual or 

inferred data in relation to reported doses and reported zeros from the dosimetry files. If the number 

of zero measurements cannot be determined from the record, determination of the missed dose 

becomes more complex. When only summary dose is known, the number of zero doses can be 

estimated based on the dose level and the monthly, quarterly, or annual limits for that year and the 

number of possible zero monitoring intervals, using the methodology in NIOSH (2006c). 

Quarterly or annual limits:

• 1951–1967. 3 rem/qtr (Figure A-10), 

• 1968–1992. 5 rem/yr (observed in Rockwell 1985), and

• 1993–2005. 2 rem/yr (DOE 1992). 
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Table 6-4 divides these dose limits into exchange frequencies. Either the dosimetry records or the 

default values from Table 6-2 should be used to determine or estimate the exchange frequency and 

number of reported zeros. Using the methodology of NIOSH (2006c), it is possible to develop an 

appropriate estimate of the number of zeros, and ultimately the missed dose, using either approach.

6.5.2 Discrepancies

If the employee’s record contains discrepancies, it is favorable to the claimant to use the higher dose 

in the dose reconstruction. Care must be taken to interpret dose numbers properly if units were not 

specified. RFP routinely used milliroentgen or millirem as the unit of dose. If a number has no unit 

indicated, it is probably not in rem (ORAUT 2018a). It is highly unlikely that a record would show a 

dose greater than the quarterly or annual limit without an additional record indicating an overexposure 

(ORAUT 2018a). 

Corrections were noted in the dose record when calculation or computer errors occurred (ORAUT 

2018a). Such corrections were usually noted on the hard-copy report, and a notation was entered if 

the electronic record was updated. If the record was updated and the update noted, the correction 

should not be applied again. If there is no obvious notation to indicate the incorporation of a 

correction, the approach more favorable to claimants is to incorporate the correction in the dose used 

for reconstruction.

6.5.3 Missing Entry

If the dosimetry history contains a missing entry, this probably indicates that the individual missed the 

dosimeter exchange and that the next dosimeter includes the dose from both exchange periods. A 

less likely indication is that the badge was lost and no dose was assigned for that period. A period of 

3 months is defined as a short gap based on the fact that the site routinely reported quarterly 

summary data. The assumption favorable to claimants is that the dosimeter was lost; dose should be 

assigned for that gap period using the following approach or coworker data: 

• If short gaps (i.e., 3 months or less for monthly monitoring or 1 quarter for quarterly monitoring) 

in the individual’s dosimetry records exist and is bounded on both ends by dosimetry data, the 

average between of individual’s adjacent monitoring data should be used to fill in the gaps in 

their dosimetry data. For 1969 and 1970, refer to Section 6.5.8 of this TBD for guidance on 

gap filling.

• If large gaps (i.e., greater than 3 months for monthly monitoring or greater than 1 quarter for 

quarterly monitoring) in the individual’s dosimetry records exist or the period is not bounded by 

dosimetry data, external coworker dose data or ambient dose data should be used to fill in the

gaps in their dosimetry data. For 1969 and 1970, refer to Section 6.5.8 of this TBD for 

guidance on the assignment of external coworker dose.

Consideration should be given to actions that might provide a basis for a gap in monitoring, such as 

the quality of the reported monitoring data, change in job category, work location, documented 

absence from work, or any administrative action.

6.5.4 Exposure Geometry

NIOSH has determined that an assumption of 100% anterior-posterior (AP) exposure for dose 

reconstructions is favorable to claimants. Glovebox workers could have experienced exposure 

geometry characteristics that could result in an underestimation of reconstructed organ doses. 

NIOSH has issued a technical information bulletin to address this issue (NIOSH 2011). 
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Table 6-4. Dose limits (rem) based on exchange frequency.

a

Year

Limit

(rem)

Period

(yr)

52

Weekly

26

Semimonthly

24

Bimonthly

12

Monthly

4

Quarterly

2

Semiannually

1951 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1952 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1953 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1954 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1955 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1956 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1957 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1958 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1959 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1960 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1961 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1962 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1963 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1964 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1965 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1966 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1967 3 0.25 0.231 0.462 0.500 1.000 3.000 Not used

1968 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1969 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1970 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1971 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1972 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1973 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1974 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1975 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1976 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1977 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1978 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1979 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1980 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1981 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1982 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1983 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1984 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1985 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1986 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1987 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1988 5 1 0.096 0.192 0.208 0.417 1.250 Not used

1989 5 1 0.096 0.192 0.208 0.417 1.250 2.500

1990 5 1 0.096 0.192 0.208 0.417 1.250 2.500

1991 5 1 0.096 0.192 0.208 0.417 1.250 2.500

1992 5 1 0.096 0.192 0.208 0.417 1.250 2.500

1993 2 1 0.038 0.077 0.083 0.167 0.500 1.000

1994 2 1 0.038 0.077 0.083 0.167 0.500 1.000

1995 2 1 0.038 0.077 0.083 0.167 0.500 1.000

1996 2 1 0.038 0.077 0.083 0.167 0.500 1.000

1997 2 1 0.038 0.077 0.083 0.167 0.500 1.000

1998 2 1 0.038 0.077 0.083 0.167 0.500 1.000

1999 2 1 0.038 0.077 0.083 0.167 0.500 1.000

2000 2 1 0.038 0.077 0.083 0.167 0.500 1.000

2001 2 1 0.038 0.077 0.083 0.167 0.500 1.000

2002 2 1 0.038 0.077 0.083 0.167 0.500 1.000

2003 2 1 0.038 0.077 0.083 0.167 0.500 1.000

a. Source: See Section 6.5.1.
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The analysis considered alternative geometry considerations as follows. Because little information is 

available on the exposure geometry for an individual, estimates have been made by the author using 

professional judgment (NIOSH 2007, Section 4.4.1) for each major job category (Attachment B). To 

estimate the exposure geometry for major job categories, engineering judgment was used and a 

simple calculation was performed. The fraction of the dose received via each geometry is a product 

of the dose rate and exposure duration that each worker experienced. Workers experienced a higher 

dose rate when working hands-on with radioactive material and a lower dose rate as they performed 

other tasks in the radiation control area. An estimate of the fraction of hands-on time was chosen for 

each major job category (ORAUT 2018a). Selection of source geometry was based on an assumed 

configuration (selected by the author) of the radioactive material to which the workers were exposed. 

From this, a relative dose was estimated for hands-on work (1 ft away) and non-hands-on work (4 ft 

away), using simple rules of thumb. These were combined to estimate the fraction of the dose 

received via the AP geometry (hands-on) or other geometries for the balance of the exposure (ICRP

1996). Table 6-5 presents these results. The non-AP exposure was estimated to come from either 

the rotational (ROT) or isotropic (ISO) geometry. The difference is that ISO geometry encompasses 

exposure from all angles (including above and below) while ROT encompasses only exposure from all 

horizontal directions to the upright individual. Chemical operators receive doses from above and 

below due to pipes in the overhead and near the floor. All others were assumed to receive their nonAP doses from the ROT geometry (ORAUT 2018a). Table 6-6 lists these fractions, which are 

rounded.

Table 6-5. Exposure geometry calculation (% dose received). 

Major job category Hands-on work (time) Source geometry AP ISO or ROT

Chemical operators 25% Line 57% 43%

Metallurgical operators 75% Point 98% 2%

Maintenance workers 75% Plane 98% 2%

Support personnel 5% Plane 46% 54%

Analytical laboratory tech. 75% Point 98% 2%

Site support personnel 0% Plane 0% 100%

Radiation control technicians 10% Plane 64% 36%

D&D workers 75% Plane 98% 2%

Table 6-6. Exposure geometry defaults for major job 

categories. 

Major job category AP ISO ROT

Chemical operators 50% 50% 0%

Metallurgical operators 100% 0% 0%

Maintenance workers 100% 0% 0%

Support personnel 50% 0% 50%

Analytical laboratory tech. 100% 0% 0%

Site support personnel 0% 0% 100%

Radiation control technicians 60% 0% 40%

D&D workers 100% 0% 0%

6.5.5 Lead Aprons

Lead aprons were available and used for a limited number of tasks at RFP. Interviews with early 

Health Physics managers indicated that they were not widely used in the early years. Lead aprons 

were used for specific tasks at different times throughout the day when operators worked in proximity 

with kilogram quantities of plutonium outside gloveboxes. When engaged in activities such as bagout 

operations and packaging and handling completed assemblies, workers often used lead aprons. 

Major job categories (see Attachment B) that were likely to use lead aprons for specific activities 

include Chemical Operators and Metallurgical Operators (including Nondestructive Testing 

Technicians). The standard procedure was to wear the dosimeter under the lead apron to measure 
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the dose to the torso (ORAUT 2018a). This does not, however, account for exposure to the 

extremities, upper arms, head, and neck. 

Available RFP external dosimetry procedures were reviewed. A June 15, 1991, procedure instructs 

workers to wear the dosimeter under the lead apron, but a March 16, 1992, draft indicates that the 

badge should be worn outside of (and taped to) the lead apron. Versions of this procedure after this 

date all support wearing the badge on the outside of the lead apron. In March 1992, a field study was 

performed in two storage vaults at RFP (Passmore 1992). This study measured Panasonic dosimeter 

response both outside and inside a lead apron fitted on a dosimetry phantom. The results of this 

study indicated that dosimeters placed under the apron detected neutrons to a significantly greater 

extent than the dosimeters placed on the outside of the apron. It is believed that the neutron albedo 

effect (low-energy neutrons reflected back into the badge) is disturbed on the outside of the lead 

apron. It is also interesting that the lead apron resulted in a reduction of less than 15% in the photon 

dose under the apron. The results of this study are applicable to dose received by workers while 

wearing aprons. Table 6-7 lists the suggested bias correction factors, which derive from the largest 

values shown in the Passmore (1992) study. The lead apron correction factors were taken as the 

maximum measured values conservatively rounded up. Thus, they represent a maximizing best 

estimate of the factor and are applied as a constant.

Table 6-7. Bias correction factors for application to dose received while wearing a lead apron. 

Cancer location Dosimeter location Neutron dose Deep photon dose Shallow photon dose

Protected area Under apron 1 1 1

Protected area Outside apron 1.9 1 1

Unprotected area Under apron 1 1.2 1

Unprotected area Outside apron 1.9 1 1

Although this field study was performed using Panasonic dosimeters, the albedo phenomenon was 

used in the Harshaw dosimeter. These bias correction factors are appropriate for application to dose 

measured by the Harshaw dosimeter while wearing a lead apron. The film and neutron track plate 

neutron dosimeters used at RFP (before 1971) did not utilize the albedo phenomenon for dosimetry. 

Therefore, it is not appropriate to use these bias correction factors for neutrons in that era. It is 

appropriate to use these factors for photon doses measured with the film dosimeters.

Adjustment to dose for use of protective lead aprons depends on the location of the cancer site in 

relation to the lead apron. The aprons covered the body from the shoulders to below the knee, but did 

not cover the arms. In later years, wraparound aprons were worn. The change in apron design has 

little effect on dose reconstruction if a 100% AP exposure is assumed (see Section 6.5.4). If the 

cancer site is under the apron, there is no adjustment (i.e., the factor is 1.0) because a dosimeter 

under the apron will reasonably measure a dose to the cancer site. If the cancer site is in an area not 

protected by an apron, and for which the dosimeter-measured dose might be too low, the 

recommended adjustment factor as listed in Table 6-7 will be applied.

If location of dosimeter is unknown, the more favorable to the claimant scenario should be assumed.

6.5.6 Recycled Uranium

Some forms of uranium metal were recycled and reprocessed within the weapons complex. There is 

a concern that workers could have been exposed to transuranic elements or fission products 

contained in these materials. Recycled uranium use at RFP was carefully reviewed and documented 

(KHC 2000a). It was determined that a very small fraction (0.03%) of the depleted uranium (DU) 

processed at RFP was known to have resulted from recycled uranium processing, and this material 

contained plutonium, neptunium, and technetium below de minimis levels.
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KHC (2000a) identified two processes that had the potential for concentrating or releasing transuranic 

elements or fission products. These processes must be considered in relation to their potential for an 

external exposure hazard that was not adequately measured by the external dosimetry used at the 

time. The two processes with potential to concentrate the recycled uranium contaminants were 

vacuum melting and casting and the “chip roaster.” 

Information from Fernald initially indicated that the vacuum melting and casting of uranium could be a 

potential concentration point. The “dross” or “skull” that formed on the top of the casting was more 

radioactive than the casting. The higher radioactivity was a result of the separation of uranium decay 

progeny (thorium and protactinium) and potentially the separation of transuranic or fission product 

contaminants. RFP did not perform analyses for these constituents at that time. Data from the 

Specific Manufacturing Capability Project at Fernald indicate that no contaminant accumulation 

occurred as a result of the melting and casting process (KHC 2000a). Even if concentration did occur, 

the external exposure potential from these contaminants would be a small fraction of the exposure 

from the concentrated uranium decay progeny and would have been adequately measured by the 

external dosimetry systems in use at the time.

The conversion of DU oxide in the RFP chip roaster in Building 444 was identified as a potential 

concentration point for recycled uranium contaminants. In this operation, turnings from machining 

activities and dross from the melting operation were converted to oxide. RFP has no direct analytical 

information on contaminant concentrations in the uranium oxide, but associated emissions monitoring 

indicates no increased levels of transuranic elements (KHC 2000a). Again, concentration of these 

contaminants would not present an external exposure hazard that would not have been adequately 

monitored by the external dosimetry systems in use at that time. 

6.5.7 Potential Elevated Background Subtraction

ORAUT-PROC-0060, Occupational Onsite Ambient Dose Reconstruction for DOE Sites (ORAUT 

2006b) identifies a concern that background dose in excess of that identified as onsite ambient 

background was removed at the time the dosimeters were processed if the background dosimeters 

received elevated exposure because of their storage in locations where background dose rates were 

high.

From the start of radiological operations at RFP in 1951 until January 1976, dosimeter background 

appears to have been determined from either laboratory blanks or control dosimeters that were stored 

on the storage boards with the dosimeters. There was some discussion that, in that period, storage 

boards might have been moved to lower dose locations because the background dose from the facility 

was unacceptably high. To validate (or dispute) this fact, a records review and interview program 

were initiated. Approximately 18 boxes of external dosimetry program records were reviewed. These 

records included weekly and monthly status reports from the 1950s, 1960s, and 1970s as well as 

some technical documents from that period. Approximately 500 pages of documents were identified 

as potentially relevant to this issue. No evidence of an identified high-background problem was found. 

No evidence of action to reduce storage-board background was found.

Interviews were conducted with [redacted] retired [redacted] managers. Each of these individuals was 

asked if they recalled this issue or actions taken in response to such a problem. None recalled 

storage-board background as a problem. Most recalled that elevated storage background was not

significant and did not affect the dosimetry results (ORAUT 2018a). 

From this review, it is concluded that elevated ambient levels of external radiation were not a problem 

at RFP during the period from 1951 to 1976.

From 1977 to February 2000, a plant-wide standard background was subtracted (ORAUT 2018a). 
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For dosimeters collected in March 2000 through 2003, badge storage-board background dosimeter 

results were used. The background dosimeter results were averaged over a 5-quarter rolling period 

and subtracted from the measured dosimeter value. An analysis of this process indicate the average 

background used was 1.14 ± 1.16 (1 standard deviation) times the previous standard background

(1977 to February 2000) (ORAUT 2018a). This information indicates that the background is in excess 

of that identified in ORAUT-TKBS-0011-4, Rocky Flats Plant – Occupational Environmental Dose

(ORAUT 2007a, Table 4-3). As indicated above, this dose was subtracted as dosimeter background 

and indicates an elevated ambient level of external radiation (ORAUT 2005a). 

6.5.8 Badge Reading Policy, 1969 to 1970

In 1969 and 1970, RFP implemented a policy of not routinely reading badges for some “low risk” 

workers. The film badges were exchanged as usual but were not read unless circumstances required 

it. A zero was then recorded in the workers record. Therefore, all reported zeros for film badges 

between 1969 and 1970, excluding gamma badges reread for the NDRP, are suspect and should not 

be used in reconstructing external dose. The worker should be treated as an unmonitored worker. 

Results reported below the LOD/2, but above zero, are considered valid, and should be used to 

assess missed dose.

Site quarterly summary data and available NDRP gamma data should be used in conjunction with 

coworker dose for addressing zeros in the dose record. Gaps in site quarterly summary dosimetry 

during this time period should not be assessed using adjacent cycle data in accordance with guidance 

in Section 6.5.3. Gaps in NDRP dosimetry during this time period can be assessed using adjacent 

cycle data in accordance with guidance in Section 6.5.3.

When assessing a quarter with site quarterly summary data, and no NDRP data, the assignment of 

coworker dose is based on guidance in Attachment A of ORAUT (2018b). When using site quarterly 

summary data, the default badge exchange frequencies for 1969 and 1970 in Table 6-2 should be 

used. 

The amount of coworker assigned, when assessing DOE quarterly data, is determined as follows:

1. A quarterly result that was originally a zero in the dosimetry records should be assigned the 

total number of coworker months in a quarter (consider the impact of partial employment 

during the quarter, if applicable). Skip the remaining steps below. Otherwise, 

2. Determine the number of badges in a quarter based on the badge exchange frequency 

(consider the impact of partial employment during the quarter, if applicable).

3. A quarterly result that is censored to zero after the LOD/2 process should be treated as a 

single badge. Proceed to Step 11.

4. Divide the quarterly summary positive results by the dose limits in Table 6-4 for the 

appropriate badge exchange frequency for 1969 and/or 1970.

5. Round up the fractional number of positive badges in Step 4 to the nearest whole number.

6. Subtract the total number of positive badges determined in Step 5 from the total number of 

badges in Step 2. Limit the number of positive badges determined in Step 8 by the number of 

badges in a quarter in Step 2.

7. Divide the quarterly summary positive results by the LOD for 1969 and/or 1970.
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8. Round up the fractional number of positive badges in Step 7 to the nearest whole number.

9. Subtract the total number of positive badges determined in Step 8 from the total number of 

badges in Step 2. Limit the number of positive badges determined in Step 8 by the number of 

badges in a quarter in Step 2.

10. Determine the median number of reported badges by averaging the results of Step 6 and 

Step 9. Do not round the number of this result.

11. Determine the number of badge exchange periods for the quarter that need coworker 

assigned by subtracting the result of Step 3 or Step 10, whichever is applicable, from the 

number of possible badges in Step 2. Do not round the number of this result.

12. Determine the amount of coworker months for a quarter by dividing the number of badge 

exchange periods determined in Step 11 by the total number of badges in a quarter and 

multiplying by the total number of months in a quarter.

6.5.9 Gammacell 220 Cobalt-60 Irradiator

There may have been more than one Gammacell 220 at Rocky Flats. The unit as designed had up to 

7 sealed cylinders, each containing up to 3,500 Ci of 60Co (RFETS 1993, pp. 4-5). Separate RFP 

reports in September 1969 (Kazanjian and Brown 1969, pp. 12, 14) and November 1971 (Lombardi 

and Donovan 1971, p. 13) describe radiolysis experiments using a Gammacell 220 containing about 

3,500 Ci of 60Co. The Rocky Flats Envision newsletter claimed in September 1999 that the 

Gammacell 220 being removed had been at RFP since late 1960s with an initial ~21,000 Ci of 60Co

(Nogues 1999, p. 2). Documentation shows the Gammacell 220 in B779, Room 218, was certified 

with an initial 60Co activity of 21,900 Ci in April 1971 (RFETS 1999c, p. 3) and weighed 8,250 lb 

(RFETS 1993, p. 4). The 60Co was contained in up to 7 cylinders, arranged radially around the active 

volume (RFETS 1993, p.5), into which samples were lowered for exposure. The sealed cylinders of 

60Co were inaccessible (Ballenger 1992, p. 8) and remained in the irradiator. The size and weight of 

the Gammacell 220 was such that it had to be lowered through a hole cut in the floor of Room 218 

into Room 121 below (RFETS 1999a, undated a, undated b). 

A former [redacted] manager responsible for the Gammacell 220 indicated that the unit was rarely 

used from 1979 until its removal in 1999 [Interview conducted on 01 Feb 2017]. The Gammacell was 

transferred to JL Shepherd in San Fernando, California after its removal from B779 (RFETS 1998b, 

1999). 

A 60Co orphan source appeared to be linked in the document with the Gammacell 220 60Co irradiator 

(RFETS 1993–2002). The 287-page document describes the orphan source in a Radiological 

Deficiency Report on the first three pages (RFETS 1993–2002, pp 2-4): 

• It was found in a cabinet in Building 779, Room 125 (November 1998);

• It was originally reported to have been disposed of in March 1995;

• There was no accompanying source registry tag; 

• No other descriptive information (activity, physical form, etc.) was provided;

• It appears finally to have been disposed of in November 1998.

The remainder of the document consisted of the engineering design package for removing the 

Gammacell 220 from Building 779 and for its disposal (RFETS 1999b). The Gammacell was located 

in Room 218 of B779 (RFETS undated a). 
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The project for removal and disposition of the Gammacell 220 began in May 1999 and was completed 

in October 1999 (RFETS 1998a, 1999b). The orphan source found in B779, Room 125, was not 

connected to the Gammacell 220 irradiator. There would have been significantly more concern than 

was expressed in the Radiological Deficiency Report if the orphan source contained even a fraction of 

the 60Co encapsulated in each of the Gammacell 220 source cylinders. The orphan source and the 

Gammacell 220 were in separate rooms in B779. The Gammacell 220 source cylinders were 

inaccessible and could not be removed for storage in a cabinet. Integrity tests and verification of the 

Gammacell 220 sources occurred before removal of the irradiator and after the orphan source was 

discovered (RFETS 1999d, p. 3). 

The Operational Health Physics Audit Report of Radioactive Sources and Radiation Producing 

Devices (Rockwell 1987) report provides source control requirements and leak test documentation for 

the Gammacell 220, for the period 1972 to 1989 and before March 1987, at RFP. It indicates that 

leak tests of 21 60Co sources, with a total activity of 22,249 Ci, showed no leaking sources (Rockwell 

1987, pp. 5, 6). A September 1992 Operational Safety Analysis of the Gammacell 220 cites hazard 

controls for elevated gamma levels and loss of containment (RFETS undated c, pp. 4-6): 

• A gamma alarm next to the source warned personnel if they need to evacuate; 

• Chirpers were worn during operations to warn of release of radioactivity;

• Complete company clothing including safety glasses was worn;

• Full-face respiratory protection was in the operator’s possession and ready for use.

An October 1992 RFP Health Physics Report specified semiannual leak tests of 60Co sealed sources 

with activity >10 μCi (Patel 1992, p. 6). A June 1993 sealed source registry safety evaluation of the 

Gammacell 220 (registry no. NR-220-D-107-S) specifies leak test intervals not to exceed six months 

using techniques capable of detecting 0.005 μCi (11,100 dpm) (RFETS 1993, p. 7) (note that this is 

the U.S. Nuclear Regulatory Commission criterion for leak testing today). A June 1999 Radioactive 

Source Report specified a semiannual leak test frequency for the Gammacell 220, noting that the last 

leak test was in January 1999 and the next was scheduled for July 1999 (RFETS 1999c, p. 3). Some 

leak test results of the Gammacell 220 are available in job specific contamination surveys from 1998 

on:

• February 1998. Results of <200 dpm/100 cm2 removable beta were reported for 15 sample 

locations on the Gammacell exterior surfaces (RFETS 1998d, pp. 3, 4); 

• August 1998. Two sample locations had <6,200 dpm/100 cm2 removable beta (RFETS 

1998c, pp. 3, 5); 

• January 1999. Three sample locations had <45.3 dpm/100 cm2 (less than the minimum 

detectable amount) removable beta (RFETS 1999d, pp. 3, 5). 

No quantitative leak test results are available during or before 1989. However, a 1987 report

indicates that of 21 sources (22,249 Ci total activity) surveyed there were no leaking 60Co sealed 

sources (Rockwell 1987, pp. 5, 6). Therefore, it is assumed that source integrity verification 

requirements existed sometime before 1989, and that some leak test measurements were conducted. 

In addition, unmonitored personnel exposure from leaking 60Co sources (including the 

Gammacell 220) are unlikely. If the Gammacell 220 had leaked, the resulting contamination would 

have been detected in later surveys, some documentation of which has been captured. Therefore, no 

additional exposures need to be assessed associated with the Gammacell 220. 
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6.6 PHOTON DOSE

6.6.1 Energy Groups

The NIOSH Interactive RadioEpidemiological Program (IREP) software for calculating the probability 

of causation (POC) (NIOSH 2007) contains three photon energy bands:

• Below 30 keV, 

• 30 to 250 keV, and

• Above 250 keV. 

Separation of the dose from each energy band is required.

6.6.1.1 Default Exposure Spectra

There is limited spectroscopy data that indicate the gamma spectrum in RFP work areas. To estimate 

the gamma spectrum to which workers were exposed, radionuclide source concentrations (DOE 

1980) for weapons-grade plutonium, enriched uranium (EU), and DU were used (freshly separated 

material) and then decayed for 10 and 30 yr. These decay times enable an understanding of the 

material to which workers were exposed. DU and EU were routinely handled in the open with no

shielding. Plutonium was almost exclusively handled in gloveboxes that provided shielding from the 

materials. The calculation assumed large pieces of material (infinitely thick in relation to the photon 

path length in that material) and 1/16-in. stainless steel as the shielding provided by the glovebox. 

Table 6-8 presents these results.

Table 6-8. Photon energy distributions. 

No shielding

Energy 

(keV)

Fresh 

EU

10-yr 

EU

30-yr 

EU

Fresh 

DU

10-yr 

DU

30-yr 

DU

<30 0% 0% 0% 0% 0% 0%

30–250 100% 99% 98% 100% 3% 3%

>250 0% 1% 2% 0% 97% 97%

1/16-in. steel shielding

Energy 

(keV)

Fresh 

Pu

10-yr 

Pu

30-yr 

Pu

<30 0% 0% 0%

30–250 100% 85% 88%

>250 0% 15% 11%

Plutonium processed at RFP has varied in age from freshly separated to wastes that have been 

stored on the site for many years. Using the default assumption that the material is freshly separated 

maximizes the dose from the 30- to 250-keV photons. Low-energy photons that are shielded in this 

analysis do, in fact, escape the glovebox through open glove ports and unleaded windows as well as 

from oxide coated on the interior surfaces of the gloves, especially when they are pulled outside the 

glovebox for storage to prevent them from being caught in machinery (DOE 2003). It has been 

estimated that approximately 25% of the dose is from <30-keV photons (ORAUT 2018a). Low-energy 

(<30-keV) photon exposure is estimated from reported penetrating and skin photon dose by use of the 

algorithms in Section 6.6.1.2.

Protactinium-234m is a decay product in the 238U (DU) decay chain and emits a 2.29-MeV beta 

particle. Therefore, a significant quantity of photons from bremsstrahlung radiation is produced and 

contributes photons of intermediate energy (30 to 250 keV). These photons are not included in 
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Table 6-8. Bremsstrahlung radiation can contribute up to 40% of the photon dose from uranium metal 

(DOE 2001). This decay product grows in fairly rapidly and is present in equilibrium quantities for 

most DU that was processed at RFP. It is appropriate to use the default assumption for DU that 50% 

of the dose is contributed by photons in the 30- to 250-keV photon energy range and 50% of the dose 

is a result of exposure from photons in the >250-keV photon energy range. 

Although EU has significantly less in-growth of 234mPa, 235U and its decay products emit 185.7-keV 

photons 57% of the time and 143.8-keV photons 11% of the time. These photons dominate the 

measured photon energy spectra. Therefore, for EU, it is appropriate to use the default assumption 

that all of the photon dose is a result of exposure in the 30- to 250-keV photon energy range. This 

assumption is favorable to claimants. The default assumptions are shown in Table 6-9. 

Table 6-9. Default photon energy distributions for gamma dose rates from field measurements.

a

Energy (keV) Plutonium EU DU

<30 25% 0% 0%

30–250 75% 100% 50%

>250 0% 0% 50%

a. In cases where the material source is unknown the most favorable to the claimant scenario, that is plausible for the 

given claim, should be applied.

Table 6-9 of the RFP External TBD indicates a default photon energy distribution of 25% <30 keV and 

75% 30-250 keV for plutonium facilities. This energy distribution refers to general photon fields, and 

would be used only if gamma dose rate data from field measurements were available instead of 

dosimetry results. The 25% <30 keV and 75% 30-250 keV photons energy split in Table 6-9 for 

plutonium workers would apply in a rare case where a worker was unmonitored and dose might need 

to be assigned using external dose rate data in the claim.

The RFP energy distributions applied for plutonium facilities for recorded dosimetry data or coworker 

dose data for dose reconstruction are shallow dose assigned as 100% <30 keV photons, and photon 

deep dose assigned as 100% 30-250 keV photons. The assumption of photon deep dose assigned 

as 100% 30–250 keV photons is claimant-favorable based on fresh plutonium from Table 6-8. The 

shallow and photon doses based on RFP dosimetry results for the appropriate eras are assigned in 

accordance with the algorithms in Section 6.6.1.2.

6.6.1.2 Dosimeter-Indicated Photon Energy

In the discussion below, a portion of the skin dose as reported in the records for plutonium facility 

workers is interpreted as exposure to low-energy photons (<30 keV) and not strictly as Hp(0.07)

(ORAUT 2018a). 

6.6.1.2.1 Pre-1960

The two-element film dosimeter used at RFP before 1960 had both OW and cadmium (Cd) filter 

components. Some dosimetry records indicate that the dose was determined from the film darkening 

under each dosimeter element, as well as the recorded skin and penetrating dose values. Based on 

review of some of these data and data after 1960, the skin dose in this era was calculated as a sum of 

each of the two windows (OW + Cd). The dosimeter could not effectively measure the 60-keV 

photons, and the reported penetrating (Pen) dose was therefore calculated from the measured OW 

and Cd component doses using the algorithm Pen = 50% • OW + Cd, and the reported Skin dose was 

calculated as Skin= OW + Cd. The addition of 50% of the recorded OW dose to the recorded Cd

dose was determined by Rocky Flats staff to correct the reported Pen dose for this underestimation 

(i.e., the relative ratio of the 60-keV photon dose to OW dose contribution was determined to be

about 0.5). 
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Therefore, the sum of the low-energy (<30 keV photons or >15 keV electrons) and the photon doses 

(30- to 250-keV and/or >250-keV photons as applicable) exceeds the original reported skin dose. As 

a result, the following method should be used to estimate the low-energy photon dose for plutonium 

workers, intermediate/high-energy photons for plutonium and uranium workers, and the >15-keV 

electrons for uranium workers:

(6-1) electrons photon Skin Pen OW > < 15keV or 30keV = [ ] ÷− 0.5 =

(6-2) photon photon Pen CD OW ( ) 30 250keV − > + = =+ ( ) 250keV 0.5

The equations above should be used with the facility radiation characteristics listed for Rocky Flats 

facilities in Tables 6-10 for plutonium, DU, or EU facilities to assign measured dose in the appropriate 

electron and photon ranges and proportions. The use of these equations might lead to a slight 

overestimate of electron dose. Alternatively, if original component data are available in the dosimetry 

record, Pen and Skin doses can be calculated using the Cd and OW results. 

6.6.1.2.2 1960 to 1970

The three-element film dosimeter used at RFP from 1960 to 1969 also had an OW, a Cd filter, and an 

additional brass (Br) filter providing half the filtration of the Cd. The Br filter was added to more 

accurately measure the 60-keV photons. This dosimeter was phased out during the year 1970.

Some dosimetry records (see Figure A-7) indicate that the dose was determined from the film 

darkening under each dosimeter element, as well as the recorded skin and penetrating dose values. 

Based on review of some of these data, the skin dose in this era was calculated as a sum of each of 

the three windows (OW + Cd + Br). The penetrating dose was calculated by adding the Cd + Br + 

35% • OW. The 35% OW addition to the deep dose was a DOE weapons complex standard practice 

during this period (including the Hanford and Savannah River Sites) to account for some low-energy 

photon (<30 keV) contribution to deep dose.

To properly reconstruct the low-energy and intermediate-energy photon dose between 1960 and 

1970, the following reverse algorithm should be applied:

(6-3) 15keV 30keV electron photon Skin Pen OW > < or = [ ] ÷− 0.65 =

( ) (6-4) photon photon Skin electron photon BR 30–250 keV 250keV – 15keV or 30keV + = > > < = + CD

The equations above should be used with the facility radiation characteristics listed for Rocky Flats 

facilities in Tables 6-10 for plutonium, DU, or EU facilities to assign measured dose in the appropriate 

electron and photon ranges and proportions. As an alternative, if original component data are 

available in the dosimetry record, Pen and Skin doses can be calculated using the Cd, Br, and OW 

results. 

6.6.1.2.3 1970 to Present

Starting in 1970, RFP used TLDs to measure photon dose, which provided substantially better 

performance than film. The TLD materials used were much more tissue-equivalent and the response 

much less energy-dependent. Dosimeters were calibrated to more appropriate photon energies, and 

filter design had advanced. Although various chip and filter combinations were used, the data in DOE 

dosimetry files do not generally include dose from individual components similar to the Cd, Br, and 

OW categories from the film era. Because doses in the era of TLD dosimetry at RFP are felt to 
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provide accurate measurements of skin and deep doses, the following relationships should be used 

for this period:

(6-5) electrons photon Skin Pen > < 15keV or 30keV = −

(6-6) photon photon Pen 30–250 keV + = >250keV

The equations above should be used with the facility radiation characteristics listed for Rocky Flats 

facilities in Tables 6-10 for plutonium, DU, or EU facilities to assign measured dose in the appropriate 

electron and photon ranges and proportions.

6.6.2 Calibration Factor

6.6.2.1 Reported-Dose-to-Organ-Dose Conversion Factor Units

Standard X-ray film was initially used for photon dosimetry at RFP. This film was processed by Los 

Alamos National Laboratory (LANL). This was followed by a period in which a subcontractor 

performed the processing, after which RFP took over the processing. 

According to ORAUT-TKBS-0010-6, Los Alamos National Laboratory – Occupational External Dose, 

the LANL dosimetry results were calibrated in roentgens (ORAUT 2013).

When RFP provided the film dosimetry, it appears that roentgens continued as the unit of calibration 

(Mann 1967). It is reasonable to assume that this continued until calibration of the Panasonic TLD 

dosimetry system, which was performed using DOE Laboratory Accreditation Program (DOELAP) 

sources at Pacific Northwest Laboratory (PNL). DOELAP sources have been used since that time. 

The personal dose equivalent [Hp(10)] is the appropriate unit to use for this period. Table 6-10 

summarizes dose units to use for organ dose conversion factors.

Table 6-10. Photon dose units for use 

with organ dose conversion factors.

a

Period Unit

1951–1982 roentgens

1982–2005 Hp(10)

a. Source: See Section 6.6.2.1.

Conversion to organ dose is accomplished using the factors provided in Appendix A of NIOSH (2007). 

Plutonium-specific photon dose conversion factors are provided in Table 4.1a of that document and 

should be applied for plutonium exposures at RFP.

6.6.3 Missed Dose

Section 2.1.2 of NIOSH (2007) recommends the use of the LOD/2 method for determining missed 

dose.

6.6.3.1 Limit of Detection

The film badge initially used at RFP is similar to that developed at the University of Chicago and used 

at other U.S. Atomic Energy Commission (AEC, a DOE predecessor agency) sites. All of these 

badges used X-ray film surrounded with a metal badge holder. They had an OW and an area covered 

with 1 mm of silver, tin, or cadmium (Alvarez et al. 2003). A PNL study of this two-element dosimeter 

(Wilson et al. 1990) identified a detection level of about 40 mR at the upper 95% confidence level for 

radium gamma radiation. Improved film, implemented at Hanford in 1960 (Wilson et al. 1990), 
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reduced this detection level to about 15 mR. Information found at RFP indicated that a DuPont 

558 film packet was used in 1964 (Mann 1964). This packet contained a DuPont 508 sensitive film 

and the insensitive DuPont 1290 film. The 1290 film was not processed unless the 508 film was too 

exposed to read. It is not clear if RFP used the earlier 502 film or, if so, when it changed to the 

508 film. Hanford changed to 508 film in 1960 (Wilson et al. 1990). It is favorable to claimants to 

assume that RFP used the less sensitive 502 film until 1960 and then used the more sensitive 

508 film. The film LOD selected is that determined by Wilson et al. (1990) for the Hanford badge.

In 1969, RFP started using Harshaw TLD chips to measure photon dose. Again, this dosimeter was 

similar to one used at Hanford. Wilson et al. (1990) identified an estimated detection level of 20 mR 

for radium gamma detection. The LOD information has not been identified specifically for TLD 

implementation at RFP, but is believed to be similar to that for the Hanford dosimeter 

(ORAUT 2018a). 

The switch at RFP to the Panasonic dosimeter in 1983 achieved improved sensitivity. Information on 

the LOD during this period has not been identified, so the value of 20 mrem, similar to that achieved in 

1982, is recommended as favorable to claimants. 

In 1992, a study was performed to reduce the variability in low-dose measurements. An uncertainty 

criterion incorporated in the algorithm resulted in more stable dose measurements at low doses. This 

resulted in an estimated LOD of 10 mrem. A dose-reporting threshold of 10 mrem was implemented. 

Any dose below this was reported as zero. Table 6-11 lists photon LODs for the RFP dosimeters.

Table 6-11. Photon LODs.

a

Period LOD

1951–1968 40 mR

1968–1982 20 mR

1983–1992 20 mrem

1993–2004 10 mrem

2005 5 mrem

a. Source: See Section 6.6.3.1.

6.6.3.2 Number of Zero Readings

Section 6.5.1 of this TBD discusses the determination of the number of zero readings.

6.6.3.3 Determination of Missed Dose

Determination of missed dose is performed using LOD/2 times the number of zero readings, as 

discussed in Section 2.1.2.2 of NIOSH (2007). For the period 1977 to 2005, the number of zero 

readings can be determined directly from the dosimetry data. The missed dose is assumed to have a 

lognormal distribution with central tendency nLOD/2, and the upper 95% dose is nLOD, where n is the 

number of zero readings. If the number of zero readings cannot be determined, it must be estimated 

under the assumption that prorated dose limits were not exceeded. Section 6.5.1 of this TBD and 

Section 2.1.2.3 of NIOSH (2007) discuss this estimate. In this case, the estimate is assumed to have 

a lognormal distribution (NIOSH 2007, Section 2.1.2.4).

6.6.3.4 Unmonitored Energy Range

All dosimeter types used at RFP were calibrated and their responses were corrected for photon 

energies that result in worker dose in the work areas (low-energy X-rays, americium photons, and 

high-energy photons). No corrections for unmonitored photon energy range are appropriate.
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Baker (2002) states that the two-element film dosimeter used at RFP was similar to those used at 

other sites. The Savannah River Site TBD (ORAUT 2005a) discusses the response of this dosimeter. 

These documents address the significant overresponse of film to low photon energies. The dosimeter 

(OW) was calibrated with low-energy photons. To correct for this overresponse, a portion of the OW 

dose was added to the deep dose measured under the 1-mm cadmium filter. There is evidence (Falk 

undated) that this correction was used at RFP. This indicates that the early film dosimeter was 

corrected for energy response. No missed photon dose correction factor is appropriate for this 

dosimetry system.

The multielement film dosimeter used at RFP provided better energy response to measure worker 

dose more accurately. Although little information is available on this dosimetry system, it appears that 

corrections were incorporated to prevent missed photon dose (Baker 2002; Putzier 1982, p. 1). 

Therefore, no missed photon dose correction factor is appropriate for this dosimetry system.

Harshaw TLD chips were used at RFP in an interim neutron film/photon TLD badge and then in the 

RFP TLD badge. These dosimeter elements were shielded and of various thicknesses. Most 

importantly, the TLD elements were relatively tissue-equivalent in relation to photon response 

(ORAUT 2006a, Section A.2.1.2) and unlikely to have missed photon dose in an energy range to 

which workers were exposed. No missed dose correction is appropriate for this dosimetry system.

The initial implementation of the Panasonic TLD system was based on a range of DOELAP exposure 

categories. The response of the dosimeter was evaluated in relation to these exposures, and the 

algorithm was derived from these exposures. Therefore, the initial implementation of the Panasonic 

TLD system and the later DOELAP-accredited operation of that system are unlikely to have missed 

photon dose in an energy range to which workers could be exposed. No missed-dose correction is 

appropriate for this dosimetry system.

6.6.4 Geometry

6.6.4.1 Angular Dependence

The film dosimeters used at RFP had varying angular responses. Dosimeters were not always 

exposed perpendicularly, which resulted in varying responses in relation to actual worker exposure. 

The film dosimeter experienced an apparent increase in dose when exposed from the edge because 

photons were able to expose the film without passing through the filter. RFP has generated limited 

experimental exposure data that demonstrate this phenomenon qualitatively. Edge-on exposure with 

60-keV photons indicated a factor of 4 overresponse.

TLD dosimeters are likely to experience the same problem. No information on this issue in relation to 

the neutron film/photon TLD badge or the Harshaw TLD badge photon response has been found 

(ORAUT 2018a). 

Quantitative information is available for the RFP Panasonic dosimeter (KHC 2001, Section 04.06.2). 

The dosimeter was tested in 1993 and 1996. For eight DOELAP exposure categories, element 

responses generally decreased as the angle increased. For angles of incidence from –30° to +30°, 

the ratio of reported dose to delivered dose ranged from 0.88 to 0.99 for photons. 

There are insufficient data to identify an angular dependence correction to apply to any of the 

dosimeters. Because any correction would reduce the dose, or in the case of the Panasonic 

dosimeter increase the dose only slightly, not including a correction factor is generally favorable to 

claimants.
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6.6.4.2 Exposure Geometry

A 100% AP exposure geometry for all external doses at Rocky Flats should be applied, in accordance 

with Section 6.5.4.

6.6.5 Uncertainty

OCAS-IG-001 describes methods for quantification of laboratory uncertainty associated with reading 

film and TLDs (NIOSH 2007). These methods provide a statistical treatment of the variability 

associated with reading dosimeters in the laboratory.

6.6.5.1 Film

RFP used film to measure photons between 1951 and 1969. The DuPont 558 film packet with the 

sensitive 508 film was used in 1964 (Mann 1964). The 508 film was the successor to 502 film, and 

each has a useful range from 10 or 20 mR up to approximately 10 R (Lalos 1989). It is not clear if 

RFP used 502 film or, if so, when it changed to 508 film. Hanford changed to 508 film in 1960 (Wilson 

et al. 1990). Both film types have approximately the same reading uncertainty.

The method in NIOSH (2007) was used to determine the laboratory uncertainty (upper 95% 

confidence dose) for film readings. This method is detailed in Film Badge Dosimetry in Atmospheric 

Nuclear Tests (Lalos 1989). The discussion of this method cites sensitivity parameters for 502 film. A 

spreadsheet was developed using these parameters to match the example provided and then 

modified with RFP-specific parameters. RFP densitometer readings appear to be a factor of 1,000 

greater than those illustrated in the example. It is believed, based on review of the records, that these 

density units are thousandths (milli-) density units. The results are consistent with the example when 

this assumption is used. A review of dosimetry worksheets indicated that density readings were 

recorded to the nearest whole number; therefore, the densitometer reading uncertainty is assumed to 

be ±0.5 density unit. A review of RFP density-to-dose conversion charts from 1966 to 1968 made a 

determination of film sensitivity possible. Using this parameter, the upper 95% confidence doses for 

various dosimeter readings were calculated. 

Although the uncertainty is lower at higher exposures, the National Research Council methodology 

recognizes that additional uncertainty contributed by variability in calibration, film processing, and 

reading the calibration curve prevents the upper 95% confidence dose from falling below 120% of the 

reported exposure. This limitation has been applied here (Table 6-12), and it affects the estimate of 

the upper 95% confidence dose above 27 mR. Table 6-13 lists uncertainties for photon film dose.

Table 6-12. Uncertainty for photon 

film dose.

Dose 

(mR)

Upper 95% confidence 

photon dose (mR)

1 6

2 7

5 10

10 15

20 25

50 60

100 120

200 240

500 600

1,000 1,200

2,000 2,400
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Table 6-13. Uncertainty for loosechip TLD photon dose.

Dose

Upper 95% confidence 

dose (mrem) 1969–1982

1 21

2 22

5 25

10 30

20 40

50 72

100 126

200 239

500 585

1,000 1,166

2,000 2,330

A default value of 1.2 can be used for the photon film dose uncertainty multiplier.

6.6.5.2 Thermoluminescent Dosimeter

TLDs provided improved photon dosimetry. This section estimates the uncertainty associated with 

this type of dosimeter for the early years of use and then discusses the measured uncertainties after 

1983 when DOELAP performance testing began.

6.6.5.2.1 Loose-Chip Thermoluminescent Dosimeters

Harshaw TLD chips were used to measure photon dose at RFP from 1969 to 1982. These chips were 

carried in a dosimeter holder but were removed to be read (thus the term loose). A calculation was 

performed to estimate the uncertainty associated with reading the photon dose from these 

dosimeters.

Little information has been found that describes the variability of response when these chips were in 

service. A chip sorting procedure was used to remove chips from service that had responded outside 

set criteria (Link and Pennock 1983). The procedure was to expose the chips to a 1,000-mrem dose 

equivalent using a 137Cs source. The chips were then read, and any that responded outside the 

±0.165 of the mean were removed from use. Assuming that the chip response had a normal 

distribution so that 5% of the chips were removed during the sorting process (an assumption favorable 

to claimants), the upper and lower cutoffs would have to be 1.96 standard deviations above/below the 

existing chip population. Therefore, the initial chip population standard deviation is (0.165 × 1,000) ÷ 

1.96 = 84.18 or 8.4%. Performing a Monte Carlo simulation on this distribution, removal of the chips 

outside the criteria results in a truncated normal distribution with a standard deviation of 7.4%. The 

higher 8.4% result was selected as a parameter that describes the chip population routinely used to 

measure dose (an assumption favorable to claimants). Using the Simplified Dosimetry Uncertainty 

calculation recommended by NIOSH (2007), and assuming the critical level is the LOD estimated in 

Section 6.3.1 of this TBD, Table 6-13 lists the upper 95% confidence doses.

If all individual dosimeter readings are over 100 mrem, then a factor of 1.26 can be used. Otherwise, 

a default value of 2.0 should be used for the photon TLD dose uncertainty multiplier.

6.6.5.2.2 Panasonic Thermoluminescent Dosimeter

Table 6-14 summarizes the uncertainty associated with DOELAP-accredited Panasonic dosimeter 

dose readings. These values were calculated using the TLD uncertainty methodology referenced in 

Section 2.1.1.3.2 of NIOSH (2007). Uncertainty is quantified in the dosimetry program documentation 
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available for a DOELAP-accredited program. The standard deviation for null readings is from a study 

performed at RFP (KHC 2001), and the relative standard deviation at high readings is the standard 

deviation of the DOELAP performance test results (KHC 2001; Stanford 1990). The reasonable worst

case values from these studies were selected to provide a result favorable to claimants. No data are 

available for the initial algorithm implementation of the Panasonic dosimetry system (1983 to 1989). 

Similar performance to that after 1990 is assumed (ORAUT 2018a). 

Table 6-14. Uncertainty for DOELAP-accredited TLD photon dose (upper 95% confidence dose in 

mrem). 

Dose

Panasonic dosimeter, 

1983–1989

DOELAP-accredited 

Panasonic dosimeter, 

1990–1998

DOELAP-accredited 

Panasonic dosimeter, 

1999–2004

1 1 1 1

2 2 2 2

5 6a 6a 6b

10 12 12 12

20 25 25 24

50 61 61 59

100 123 123 118

200 245 245 235

500 614 614 588

1,000 1,227 1,227 1,176

2,000 2,455 2,455 2,353

a. 1.23 multiplier for any dose greater than 2 mrem.

b. 1.18 multiplier for any dose greater than 2 mrem.

A default value of 1.23 for 1983 to 1998 and 1.18 for 1999 to 2004 can be used for the photon TLD 

dose uncertainty multiplier.

6.7 NEUTRON DOSE

6.7.1 Energy Groups

The measured neutron dose must be divided into energy groups consistent with the dose conversion 

factors provided in Appendix B of NIOSH (2007). These energy groups and the associated radiation 

weighting factors (wR) from International Commission on Radiological Protection (ICRP) 

Publication 60 (ICRP 1991) are:

• <0.01 MeV (wR = 5), 

• 0.01 to 0.1 MeV (wR = 10), 

• 0.1 to 2 MeV (wR = 20), 

• 2 to 20 MeV (wR = 10), and

• >20 MeV (wR = 5). 

The analysis in this section is based on neutron spectra measured at RFP (Brackenbush et al. 1989).

6.7.1.1 Exposure Spectra

In August and September 1988, PNL provided technical assistance to RFP for neutron and photon 

dose measurements (Brackenbush et al. 1989). This activity performed multisphere neutron 

measurements in representative high-neutron dose situations. The measurements included 

production locations, mockup situations in which plutonium parts were in a glovebox where 

measurements could be performed, and waste storage locations. Neutron shielding similar to that 
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experienced by workers in that area was in place. Relatively long (several-day) measurements were 

required to acquire sufficient dose to achieve accurate results.

The neutron spectra were determined from the multisphere measurements and presented in the PNL 

report. Dose rate was derived from neutron flux density information and flux-to-dose conversion 

factors from National Council on Radiation Protection and Measurements (NCRP) Report 38 

(NCRP 1971). No neutron flux was identified for energies greater than 20 MeV. For this TBD, the 

dose rate information was divided into energy groups as required for NIOSH dose reconstruction. 

Table 6-15 lists this information.

Table 6-15. Neutron dose measurements divided into energy groups (fraction of dose per energy 

group).

a

Location

Dose 

rate 

(rem/hr)

Avg. 

energy 

(MeV)

<10 

keV

10–

100 keV

0.10–

2 MeV

2–

20 MeV

>20 

MeV

Building 771 fluorinator line 6.07E-04 0.33 0.090 0.028 0.678 0.204 0.000

Building 771 Tank 554 4.65E-03 0.91 0.025 0.014 0.600 0.361 0.000

Building 776 molten salt glovebox 1.71E-03 0.45 0.038 0.023 0.840 0.099 0.000

Building 776 molten salt storage vault 8.84E-03 0.39 0.085 0.015 0.711 0.189 0.000

Building 776 drum storage 2.46E-02 0.63 0.027 0.034 0.689 0.250 0.000

Building 707 high dose pit 7.35E-04 1.96 0.006 0.006 0.437 0.552 0.000

Building 707 low dose pit 2.88E-04 0.92 0.015 0.009 0.758 0.218 0.000

Building 707 oxide can 1.43E-03 0.85 0.018 0.019 0.676 0.286 0.000

Building 707 plutonium ingot 1.98E-03 1.00 0.014 0.002 0.791 0.193 0.000

Mean N/A N/A 0.035 0.017 0.687 0.261 N/A

Standard deviation N/A N/A 0.031 0.010 0.117 0.130 N/A

a. N/A = not applicable.

6.7.1.2 Reported Dose to Energy Groups

This information does not show a clear pattern. Therefore, it is appropriate to apportion dose based 

on the mean breakdown listed in Table 6-15. Table 6-16 lists the default values selected from 

Table 6-15 for dose reconstruction (ORAUT 2018a). 

Table 6-16. Default neutron energy distribution. 

Neutron energy 

intervals

Fraction of dose 

(NCRP 38)

Dose multiplier 

(ICRP 60) Dose multipliera

<10 keV 0.035 2.13 0.0755

10–100 keV 0.017 1.86 0.0309

0.1–2 MeV 0.687 1.91 1.31

2.0–20 MeV 0.261 1.32 0.345

>20 MeV 0 None None

a. Multiply the reported dose by these factors to determine the ICRP 60 neutron dose for 

each neutron energy interval.

The doses and fractions discussed above are based on quality factors published in NCRP (1971). 

NIOSH (2007) indicates the use of wRs from ICRP Publication 60 (ICRP 1991). To perform this 

correction, the neutron energy deposition values in rad for each energy were multiplied by the ICRP 

wR to determine the corrected dose equivalent. These values were totaled for each neutron energy 

interval used in this dose reconstruction and compared with the value determined previously using 

quality factors from NCRP (1971). Column 3 of Table 6-16 lists the multipliers that were determined 

for each neutron energy interval. The fraction of the dose using NCRP (1971) quality factors and the 

dose multiplier using ICRP (1991) wRs were combined to determine a dose multiplier (column 4 of 

Table 6-16). The neutron dose reported in the worker's dose record should be multiplied by these 

factors to determine the ICRP (1991) neutron dose for each neutron energy interval.
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6.7.2 Calibration Factor

6.7.2.1 Dosimeter-Specific Quality Factor Conversion

The correction factors to convert from NCRP (1971) quality factors used in the neutron spectra 

measurements and the ICRP (1991) wRs are discussed in Section 6.7.1.2 and listed in Table 6-16. 

Conversion to organ dose is accomplished using the factors provided in Appendix A of NIOSH (2007).

6.7.2.2 Reported-Dose-to-Organ-Dose Conversion Factor Units

RFP initially used neutron track plates. These dosimetry elements were provided and processed by 

LANL. DDE is the unit determined to be appropriate (ORAUT 2013).

Neutron film was initially calibrated with an apparently unmoderated polonium-beryllium source. In 

1962 or 1963, this was changed to plutonium fluoride (PuF4) (Mann and Boss 1963). The dose rate 

assigned to the source was the total dose for an energy of 1.4 MeV from the National Bureau of 

Standards Handbook 63 (NBS 1957). A set of polyethylene moderators was constructed. The 

spectra from these moderated sources compared well with work area spectra measured with a 

precision long counter and a series of paraffin moderators fitted over the counter (Mann and Boss 

1963). Ambient dose equivalent [H*(10)] is appropriate for this dosimeter.

Harshaw TLDs at RFP were initially calibrated using a 210-g PuF4 source built at RFP and calibrated 

at the LANL standard pile, which was established as a neutron flux standard (Mann and Boss 1963). 

A set of polyethylene moderators was constructed to provide various degrees of moderation. The 

bare PuF4 source dose rate was calculated using neutron spectra from an unknown reference 

document and quality factors published in DOE Orders (Falk 1975). The dose rates for the 

moderated spectra were measured with currently available neutron dose rate instrumentation. The 

PuF4 source was placed in storage in about 1975 and replaced with a commercially manufactured and 

calibrated 252Cf source. The calculation of the dose rate used a published spectrum and dose rate 

(Barker 1969). A set of polyethylene moderators was manufactured for this source and ambient dose 

equivalent rates were determined in a manner similar to that used for the PuF4 source. Therefore, the 

ambient dose equivalent [H*(10)] is the appropriate unit for this period.

Panasonic TLDs at RFP were calibrated with DOELAP exposure standards. In the early 1980s, PNL 

was developing the neutron standards that were used for the original DOELAP performance testing. 

The development of all Panasonic dosimeter algorithms used at RFP was based primarily on these 

exposures. Therefore, the DDE [Hp,slab(10)] is appropriate.

Table 6-17 summarizes the dose units to use for organ dose conversion factors.

Table 6-17. Neutron dose units for use 

with organ dose conversion factors. 

Period Unit

1951–1983 H*(10)

1983–2005 Hp,slab(10)

6.7.3 Missed Dose

6.7.3.1 Limit of Detection

LANL processed neutron track plates for RFP from 1951 to 1956. The performance of this system is 

documented in ORAUT (2013). The minimum detectable dose is identified as <50 mrem.
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In 1957, RFP switched to NTA film that was processed and read by a subcontractor. Little is known 

about this processing period, so again an LOD of 50 mrem is assumed.

Beginning in July 1958, RFP processed NTA film at the site. The NDRP (Falk et al. 2005) assembled 

a processing history that is summarized in Table 6-18. Based on a background (blank) reading of 

16 tracks per 10 mm2 reported by Mann and Boss (1963) for 1962, LODs were calculated based on 

the most conservative calibration factor.

Table 6-18. Neutron film track counting detail.

a

Date Determined positive

Calibration

(mrem/track/mm2) LOD (mrem)

1959 >2 × blank 40 128

1960 >2 × blank 40 128

1961 >1.5 × blank 40 96

1962 >blank + 1.65 × sqrt(blank) 40 or 100 226

1963 >blank + 1.65 × sqrt(blank) 100 226

1964 >2 × blank 100 or 70 320

1965 >2 × blank or all 70 or 40 224

1966 All 110 Not determined

a. sqrt = square root.

Mann and Boss (1963) determined that a typical background film for 2 weeks had 16 tracks per 

10 mm2

. Using three times the standard deviation of the background and a 10-mrem/track calibration 

factor, the minimum detectable dose is 120 mrem. 

Based on the LOD, the value most favorable to claimants was selected for each year. The estimates 

from Mann and Boss (1963) were used for years when LODs were not used or not known.

In 1971, RFP started using an albedo neutron TLD. Documentation of the research performed to 

develop this dosimeter (Falk 1971) indicates a practical lower neutron dose limit of 10 to 20 mrem in 

the presence of a photon dose as high as 100 mrem. The upper limit of this estimate was selected as 

the LOD for this dosimeter.

In 1983, the Panasonic UD-809 dosimeter was introduced at RFP to measure neutrons. Data are not 

available on the LOD for this dosimeter system. Because the hardware is the same as that used in 

1990, it was assumed to be similar to performance of the system at that time. The assumed LOD is 

32 mrem.

In 1990, an algorithm update was incorporated in the Panasonic dosimetry system (Stanford 1990). 

The documentation cites a minimum detectable neutron dose of 15 to 32 mrem for a moderated 252Cf 

source.

In 1993, an algorithm update was incorporated in the Panasonic dosimetry system (KHC 2001) to 

include element reading uncertainty controls to reduce large dose fluctuations at low dose. This 

update, which has passed DOELAP performance testing, results in a stated minimum response for 

routine RFP neutron fields of approximately 15 mrem. Table 6-19 includes this value.

6.7.3.2 Number of Zero Readings

Section 6.5.1 of this TBD discusses the number of zero readings.
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Table 6-19. Neutron LODs (mrem). 

Period LOD

1951–1958 50

1959–1960 128

1961 120

1962–1963 226

1964 320

1965 224

1966–1970 120

1971–1982 20

1983–1992 32

1993–2004 15

2005 20

6.7.3.3 Neutron Dose Reconstruction Project

In the early 1990s, RFP addressed the issue of neutron film processing. It had been long recognized 

that, in the dosimetry laboratory, human factors associated with reading large numbers of neutron 

films under a microscope can significantly affect neutron dosimetry results. A pilot study in 1994 

reevaluated neutron doses for selected plutonium workers. This study indicated that the original 

evaluations of films might have contained significant errors, and that the resultant neutron doses 

might be significantly higher or lower than the doses actually received. The NDRP was initiated to 

provide current and former radiation workers an assessment of the neutron exposure received in the 

plutonium production facilities. The scope of this project covered 1952 to 1970. 

Two methods were used to identify workers for evaluation by the NDRP. The initial method was 

identification of workers using the neutron dosimetry worksheets. These sheets identified those 

workers assigned neutron-sensitive elements (i.e., neutron films or glass plates). A portion of the 

neutron worksheets indicates issue of neutron dosimeters to personnel whose home building 

assignment was not a plutonium production building (such as Buildings 21, 22, 23, 34, 44, 81, and 

86). These individuals worked in non-neutron buildings but were issued neutron dosimeters because 

they occasionally performed work activities in plutonium production buildings. Examples of these job 

descriptions include guards, radiation monitors, technical researchers, and uranium process 

operators.

The second identification method was through use of the beta-gamma worksheets for plutonium 

production buildings. The analysis used only the beta-gamma worksheets from the plutonium 

production buildings (any building with a number starting with 7), Buildings 91 and 86, and the 

combined worksheets for Buildings 21, 22, and 23. The rosters from the beta-gamma worksheets for 

these buildings were used to identify workers. Beta-gamma worksheets for other buildings were not 

used. 

The NDRP data are in three main types of image files. The NDRP protocol document contains 

detailed descriptions of the content of these files (Falk et al. 2005).

• Neutron Dose Summary. Yearly summary of the neutron dose components, the errors 

associated with the values, and reconstructed gamma dose. 

• Neutron Dose Detail. Details, by exchange period in a given year, of the neutron dose 

components. 

• Individual Timeline. Further details by exchange period in a given year about the original 

neutron dose, calculation of NDRP dose, building location, and penetrating gamma dose. 
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The total neutron dose needed for dose reconstruction consists of three quantities that are listed in 

the summary and detail presentations. These quantities are:

• Original neutron dose. The original neutron dose equivalent that was read from the film or 

glass plate. 

• Non-affected original neutron dose. The original neutron dose equivalent that was not affected 

by an NDRP dose based on a reread film or glass plate. A blank in this field indicates either 

that the original neutron dose was affected or that there was no original neutron dose on the 

dosimetry worksheet for a given monitoring period.

• NDRP neutron dose. The neutron dose estimate from the NDRP evaluation (from reread films 

or glass plates). A blank in this field indicates either that the original neutron dose was verified 

to be correct or that the original film or plates were unreadable.

• Notional neutron dose. The neutron dose equivalent estimated by the NDRP evaluation for

each gap. Notional dose is a weighted combination of two dose determination methods. 

Method 1 is based on the worker’s average neutron dose per day obtained from films 

reevaluated by the NDRP for a given calendar year and building. Method 2 is based on

applying an average neutron-to-gamma ratio to reported penetrating gamma doses for a given 

building and calendar year. Section 11.3 of Falk et al. (2005) contains a comprehensive 

description of the notional dose calculation. A blank in this field indicates that the original dose 

report was correct or that the reread film or plate (dose under the “NDRP Neutron Dose” 

heading) was adequate to replace the original film, or that there was no indication of neutron 

exposure based on the buildings the employee was in and that there was no gamma dose that 

could be used to estimate the exposure.

The Neutron Dose Summary file provides error values for the annual doses in the form of standard 

deviations (1 sigma) for the NDRP and notional doses. The NDRP and notional doses have normal 

distributions with 1-sigma error values. The non-affected dose also has a normal distribution, but 

dose reconstructors must calculate the error associated with this component in accordance with 

NIOSH (2007).

Only a limited portion of the NDRP neutron dose components can be used in dose reconstructions:

• 1952 through 1966. Only reported non-affected original neutron dose and NDRP neutron 

dose should be used in the reconstruction. Original and notional doses should not be used in 

the reconstruction of neutron doses. During periods where only original and notional doses

are reported, the worker should be treated as an unmonitored worker. Unmonitored neutron 

dose cannot be reconstructed during this period. However, where only original and notional 

doses are reported for a badge and there is adequate adjacent NDRP or non-affected original 

neutron data for a claim, these neutron dose short term gaps (i.e., 3 months or less) can be 

filled in according to NIOSH 2007. The dose estimate is still considered a partial estimate.

• 1967 through 1970. Only reported non-affected original neutron dose and NDRP neutron 

dose should be used in the reconstruction. Original and notional doses should not be used in 

the reconstruction of neutron doses. During periods where only original and notional doses 

are reported, the worker should be treated as an unmonitored worker. Original and notional 

neutron dose short term gaps (i.e., 3 months or less) may be filled in based on adjacent cycle 

non-affected original neutron dose and NDRP data. Coworker data should be used in 

reconstructing unmonitored neutron dose during this period for larger gaps in the neutron 

dosimetry data. 
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The NDRP also reported gamma doses. These doses only include those gamma doses that were in 

some way used to estimate neutron dose and, therefore, are usually less than the DOE dosimetry 

reports from the site. In some cases, the gamma dose in the NDRP evaluation files is greater than 

the penetrating gamma dose reported in the DOE site dosimetry files. In that event, the greater dose 

should be used for maximizing and best-estimate cases. The smaller dose may be used for 

minimizing cases.

For workers who had wrist dosimetry but no whole-body dosimetry, the NDRP project calculated a 

reconstructed gamma dose based on a ratio of wrist to whole-body dose. Both the Summary file and 

the gamma section of the Individual Timeline file list this dose denoted by an asterisk (*). Because 

this gamma dose does not appear in the original DOE data, dose reconstructors should add it to the 

reconstruction. 

The Individual Timeline files provide gamma data for each readout cycle that was needed to estimate 

neutron dose. As discussed above, the gamma data are not necessarily complete for any year 

(including 1970), because that was not the purpose of the NDRP evaluation and so the data cannot 

be used to directly estimate the number of recorded zeros. However, the reported positive results can 

be subtracted from the maximum number of likely zeros based on this TBD; this would represent the 

maximum number of zeros possible. The minimum number of zeros can be estimated by directly 

counting reported zeros in the Individual Timeline file. A best estimate of zeros can be determined 

using the original site quarterly DOE data and the approach from ORAUT (2018b). These data will 

allow more accurate estimation of the number of recorded zero doses for purposes of calculating the 

missed gamma dose using the LOD/2 method. The missed zeros using original site quarterly DOE 

data and the approach from ORAUT (2018b) can be compared with the number of zeros based on the 

NDRP data, and the higher of the two should be assigned.

Blank NDRP gamma badges through 1968 are considered zeros. Because missed dose is 

determined based on site quarterly data in the DOE data using the approach in ORAUT (2018b), any 

comparison of zeros with the NDRP gamma would not underestimate the missed dose assigned. 

This is reasonable because all the reported positive dose is accounted for in the comparison between 

the original reported DOE doses and the NDRP gamma doses, and assigning coworker dose for 

blank NDRP badges for quarters with DOE data would overestimate the external gamma dose. 

6.7.3.4 Default Neutron-to-Gamma Ratio

The Oak Ridge Associated Universities (ORAU) Team developed geometric mean (GM) and 

geometric standard deviation (GSD) neutron-to-photon ratios from the available RFP worker files and 

supervisor dosimetry reports. Tables 6-20 and 6-21 list the values for the periods from 1970 to 1976 

and from 1977 to 2000, respectively.

6.7.4 Geometry

6.7.4.1 Angular Dependence

Film neutron dosimeters generally record a slightly increased dose when exposed from forward 

angles other than perpendicular. It is favorable to claimants to ignore this slight difference.

The Panasonic dosimeter was evaluated for angular dependence. For neutron fields, the element 

responses generally decreased as the angle between the incident radiation and the plane 

perpendicular to the TLD increased. For angles of incidence from –30° to +30°, the ratio of reported 

dose to delivered dose ranged from 0.87 to 1 for neutrons. This slight variability does not warrant a 

specific correction.
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Table 6-20. RFP lognormal neutron-to-photon ratio values, 1970 to 1976.

a

Year

Neutron-to-photon ratio GM 

(semimonthly exchange) GSD

95th 

percentile

Neutron-to-photon ratio 

GM (monthly exchange) GSD

95th 

percentile

1970 1.61a 3.45 12.4 N/A N/A N/A

1971 1.61 3.45 12.4 N/A N/A N/A

1972 1.32 2.15 4.64 0.8 2.63 3.94

1973 1.32c 2.15 4.64 N/A N/A N/A

1974 0.68 3.01 4.16 N/A N/A N/A

1975 0.67 3.31 4.82 N/A N/A N/A

1976 0.95 3.59 7.81 0.78 4.29 8.55

a. N/A = not applicable.

b. Data for 1970 were not available. This value is the greater of the ratios for 1969 and 1971. The high neutron-to-gamma 

ratio in 1971 for Building 771 is reasonable from a process aspect, and extrapolating that ratio back to 1970 also is 

reasonable. In the aftermath of the 1969 plutonium fire in Buildings 776 and 777, the salvaged plutonium oxide had to 

be converted back to plutonium metal, a process done in Building 771. Building 771 had a huge backlog for relatively 

pure plutonium (little americium, low gamma) to be reprocessed, which was staged in the 776 to 771 tunnel and any 

other possible staging area in the vicinity. In addition, PuF4 (a high neutron source) seemed to be generated faster in 

Building 771 than it could be reduced to metal, which also caused a staging problem in or near the process areas in 

Building 771 until the backlog could be reduced.

c. Data for 1973 were not available. This value is the greater of the ratios for 1972 and 1974.

Table 6-21. ORAU Team-developed neutron-to-gamma ratios. 

Year GM GSD

1977 0.33 3.31

1978 0.57 2.53

1979 0.37 3.31

1980 0.43 2.57

1981 0.51 2.41

1982 0.4 2.4

1983 0.42 3.01

1984 0.41 3.07

1985 0.42 3.37

1986 0.49 2.93

1987 0.36 3.56

1988 0.6 2.83

1989 0.36 5.5

Year GM GSD

1990 0.35 3.77

1991 0.29 2.41

1992 0.4 2.03

1993 0.61 1.93

1994 0.41 2.35

1995 0.35 2.28

1996 0.26 4.1

1997 0.44 2.18

1998 0.39 3.03

1999 0.45 2.79

2000 0.6 3

Overall 0.42 3

6.7.4.2 Exposure Geometry

A 100% AP exposure geometry for all external doses at Rocky Flats should be applied, in accordance 

with Section 6.5.4.

6.7.5 Uncertainty

6.7.5.1 Film

The NDRP has evaluated film uncertainty. For this reason, uncertainty of film neutron dosimeters is 

not discussed in this TBD. 

6.7.5.2 Thermoluminescent Dosimeter

Falk (1971) describes the Harshaw TLD system development. That document describes field tests in 

RFP plutonium production facilities. The results indicate “that the survey dose range is consistently 

within 20 percent of the TLD neutron dose indication.” Therefore, for the Harshaw neutron dosimeter, 

a 95% confidence interval of 20% has been selected, and the standard deviation is 20% ÷ 1.96 = 

10.2%. The methodology for TLD uncertainty in NIOSH (2007) is used.
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The initial Panasonic TLD algorithm was evaluated during development (Author unknown undated). 

The results of the evaluation stated, “A large number of (relative) biases in the range –0.100 to +0.100 

and the paucity of the (relative) biases outside the ±0.200 range indicate a robust, effective algorithm.” 

Based on this evaluation, the maximum relative bias of 0.206 was selected as the 95% confidence 

interval, and a standard deviation of 0.206 ÷ 1.96 = 10.5% was thereby determined.

The Stanford (1990) algorithm upgrade was tested during DOELAP performance testing. The 

unmoderated neutron dose category resulted in a standard deviation of 0.072. This value was 

selected to determine the upper 95% confidence dose during this period.

The 1993 algorithm upgrade (KHC 2001) was tested during 1999 DOELAP performance testing. The 

unmoderated neutron dose category resulted in a standard deviation of 0.065. A mixture of neutrons 

with both low- and high-energy photons was tested. The worst case standard deviation was 0.09. 

This value was selected to determine the upper 95% confidence dose for all dates after the 

implementation of this algorithm.

Table 6-22 lists the uncertainties for these dosimetry systems.

Table 6-22. Uncertainty for TLD neutron dose measurements (upper 95th% confidence dose in 

mrem). 

Dose 

(mrem)

Harshaw TLD 

dosimeter, 

1971–1982

Panasonic dosimeter, 

1983–1990

DOELAP-accredited 

Panasonic dosimeter, 

1991–1992

DOELAP-accredited 

Panasonic dosimeter, 

1993–2004

1 1.25 1.25 1.21 1.23

2 2.43 2.44 2.32 2.38

5 6.01a 6.04b 5.72c 5.89d

10 12 12 11 12

20 24 24 23 24

50 60 60 57 59

100 120 121 114 118

200 240 241 228 235

500 600 603 571 588

1,000 1,200 1,206 1,141 1,176

2,000 2,400 2,412 2,282 2,353

a. 1.20 multiplier for 5 mrem or greater.

b. 1.21 multiplier for 5 mrem or greater.

c. 1.14 multiplier for 5 mrem or greater.

d. 1.18 multiplier for 5 mrem or greater.

Default values of 1.2 for 1971 to 1982, 1.21 for 1983 to 1990, 1.14 for 1991 to 1992, and 1.18 for 

1993 to 2004 can be used for the neutron TLD dose uncertainty multiplier.

6.8 ELECTRON DOSE

Beta radiation fields are usually the dominant external radiation hazard in facilities that require contact 

work with unshielded forms of uranium. This was the case at RFP for EU and DU work. It should be 

assumed that the skin dose reported in RFP dosimetry records for uranium workers results from 

electrons with energies greater than 15 keV.

Figure 6-9 shows estimated beta dose rates from a semi-infinite slab of uranium metal at various 

enrichment levels. For uranium enrichments up to 30%, the beta radiation field is dominated by 

contributions from 238U decay products. For DU, therefore, dose involves essentially the 2.29-MeV 

(maximum) beta particles from 234mPa, the most energetic contributor to the beta exposure.
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Figure 6-9. Estimated beta dose rate from uranium metal at various enrichment 

levels (DOE 2001).

Processes that separate and sometimes concentrate beta-emitting uranium daughters are not 

uncommon in DOE uranium facilities. The uranium foundry operations at RFP produced skull that 

resulted in high beta dose rates. Surface beta dose rates on the order of 1 to 20 rad/hr have been 

observed at some DOE facilities. Exposure control is complicated by the fact that considerable 

contact work takes place in facilities that process uranium metal. At RFP, large foundry ingots were 

generally handled by lifting devices, but machined uranium parts were handled with gloved hands. 

RFP did have problems with elevated beta dose rates from contamination on leather gloves worn 

during foundry operations (ORAUT 2018a). 

6.8.1 Energy Groups

6.8.1.1 Exposure Spectra

The beta spectrum from uranium is highly dependent on the quantity of progeny in the uranium, which 

in turn is dependent on the enrichment level of the uranium. DU progeny grow into secular 

equilibrium relatively quickly (~30 days) and can be conservatively assumed to be present at these 

levels. 

Figure 6-10 shows the relative dose rate in relation to energy. DU would be similar to the natural 

uranium used for this experiment.

6.8.1.2 Reported Dose to Energy Groups

NIOSH (2007) indicates that because extensive research in the areas of dosimeter wear location, 

electron energy spectra, and film response is required to convert dose readings to shallow dose 

properly, “... the exposure is assumed to be equal to the shallow dose [Hp(0.07)], recognizing that this 

is an overestimation of the true shallow dose. Until further research is conducted, this assumption is 

considered reasonable.” This assumption is favorable to claimants for RFP. Additional guidance for 

calculating shallow dose can be found in ORAUT-OTIB-0017, Interpretation of Dosimetry Data For 

Assignment of Shallow Dose (ORAUT 2005b).
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Figure 6-10. Shallow dose rate from natural uranium slab (DOE 2001). 

6.8.2 Calibration Factor

6.8.2.1 Reported-Dose-to-Organ-Dose Conversion Factor Units

Film dosimeters at RFP appear to have been calibrated in contact with uranium slabs. Although RFP 

documents in the 1960s report the dose rate from a uranium slab as 240 mR/hr, 240 mrad/hr, and 

240 mrem/hr at the surface, it is assumed these were inaccurate references to a dose rate in millirad 

per hour. The wR for electrons at all energies is 1 (ICRP 1991); therefore, reported beta doses are 

equivalent to rem. This value is used directly for the Hp(0.07) dose.

6.8.3 Missed Dose

6.8.3.1 Limit of Detection

Beta dosimetry at RFP used OW film calibrated to a uranium slab. ORAUT (2007b, Section 6.5.2) 

states that the minimum detectable beta dose would have been similar to that for photons. Therefore, 

40 mrem was selected as the minimum detectable beta dose appropriate for the film dosimetry period.

Harshaw TLDs were used for beta detection starting in 1969. ORAUT (2007b, Section 6.5.2) states 

that the minimum detectable dose would have been similar to that for photons. Wilson et al. (1990) 

determined that the Hanford TLD system had a 20-mR minimum detectable dose. RFP TLD 

measurements were similar. A minimum detectable dose of 20 mrem beta (shallow) is appropriate for 

RFP during this period.

The algorithm initially developed for Panasonic TLD system implementation in 1983 contains a 

constraint to ensure that the shallow dose equivalent does not fall below 0.9 times the deep dose from 

photons. Therefore, the shallow minimum detectable dose is 0.9 times that determined for deep dose 

photons (20 mrem) for this system. The minimum detectable shallow dose for this period was 

determined to be 20 × 0.9 = 18 mrem (shallow) as indicated in Table 6-23. 
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Table 6-23. Beta LODs (mrem). 

Period LOD

1951–1968 40

1969–1982 20

1983–1989 18

1990–1992 80

1993–2004 15

2005 5

In 1990, the algorithm for the Panasonic dosimetry system was improved. The documentation for this 

algorithm cites “... a minimum reportable beta dose of 25% of the total shallow dose (photon plus 

beta) or approximately 80 mrem for DU...” (Stanford 1990). It also states that “... beta doses 

delivered to radiation workers in the plant environments will likely be overestimated.” A decision to 

use the maximum 80-mrem (shallow) minimum detectable dose was made to be favorable to 

claimants. This is a significant increase in the minimum detectable beta dose. A review of the 

algorithm documentation (Stanford 1990) indicated that a constraint was incorporated into the 

algorithm to report beta dose only if the net OW (element 1) value was over 25 mR (137Cs exposure 

response). This net element reading is determined by subtracting the expected photon response and 

the expected neutron response for that element, as determined by the relationship with other 

dosimeter elements in the badge. These calculations would result in significant variability in the net 

element 1 response, and it is assumed that the constraint was included to reduce the variability in the 

resultant beta dose estimate to an acceptable level. The result is a significantly higher minimum 

detectable dose, however. This constraint appears to have been removed in the next algorithm 

update.

In 1993, an algorithm update was incorporated in the Panasonic dosimetry system (KHC 2001) to 

include element-reading uncertainty controls to reduce large dose fluctuations at low dose. This 

update has passed DOELAP performance testing and results in a stated minimum response for 

routine RFP beta fields of approximately 15 mrem (shallow) (Author unknown 1993). This value has 

been incorporated in Table 6-23. 

6.8.3.2 Number of Zero Readings

The number of zero readings is determined as discussed in Section 6.5.1 of this document.

6.8.3.3 Unmonitored Energy Range

Film and TLD are believed to respond to beta energies of dosimetric importance (ORAUT 2018a). 

There is therefore no unmonitored energy range for which a correction factor is appropriate.

6.8.4 Geometry

6.8.4.1 Angular Dependence

The sensitive dosimeter elements are mounted in a dosimetry badge. The assembled badge displays 

severe angular dependence to beta exposure, but in most cases normal worker movement tends to 

average out some of this dependence (DOE 2001).

For beta fields, the element responses of the Panasonic dosimeter generally decreased as the angle 

between the incident radiation and the plane perpendicular to the TLD increased from 0°. For angles 

of incidence from –30° to +30°, the ratio of reported dose to delivered dose ranged from 0.36 to 0.59 

for beta particles (KHC 2001, Section 04.06.2). However, based on the averaging effect cited in 

DOE (2001), no angular correction factor is proposed.
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6.8.4.2 Exposure Geometry

Exposure geometry is not a significant issue with skin exposure. Nonpenetrating radiations do not 

significantly expose tissue in other than perpendicular exposures. 

6.8.5 Uncertainty

The method in NIOSH (2007) was used to determine the uncertainty (upper 95% confidence dose) for 

film readings. This method is based on a statistical discussion in Film Badge Dosimetry in 

Atmospheric Nuclear Tests (Lalos 1989).

6.8.5.1 Film

RFP used film to measure beta dose between 1951 and 1968. This is the same film described in 

Sections 6.3.1 and 6.5.1 of this TBD. The method in NIOSH (2007) was used to determine the 

laboratory uncertainty (upper 95% confidence dose) for film readings. This method is detailed in Film 

Badge Dosimetry in Atmospheric Nuclear Tests (Lalos 1989). This TBD analysis used a similar 

uncertainty estimation methodology and developed a spreadsheet that matched the illustration given 

in Lalos (1989). A review of RFP density-to-beta dose conversion charts from 1966 to 1968 

determined film sensitivity. A saturation density for DuPont 502 film was assumed. Using this 

approach, the upper 95% confidence doses for various beta doses were calculated. A limit of 120% 

was applied as discussed in Section 6.6.5.1. This limit affects the upper 95% confidence dose at 

77 mrad and above. Table 6-24 lists these upper 95% confidence doses.

Table 6-24. Uncertainty for beta film readings.

Dose (mrad) Upper 95% confidence dose (mrad)

1 17

2 18

5 21

10 26

20 36

50 66

100 120

200 240

500 600

1,000 1,200

2,000 2,400

6.8.5.2 Thermoluminescent Dosimeter

TLDs provided improved beta dosimetry. Harshaw TLD chips were used to measure beta dose at 

RFP from 1969 to 1982. This section estimates the uncertainty associated with this type of dosimeter 

for the early years of use and then discusses the measured uncertainty when DOELAP performance 

testing was initiated.

6.8.5.2.1 Loose-Chip Thermoluminescent Dosimeters

Harshaw TLD chips were used to measure beta dose in parallel with photon dose. As with the photon 

TLD uncertainty, the chip-sorting procedure was used to estimate the standard error associated with 

the beta TLD measurements. Using the Simplified Dosimetry Uncertainty calculation recommended 

by NIOSH (2007), and assuming that the critical level is the beta LOD estimated in Section 6.8.3.1 of 

this TBD, Table 6-25 lists the upper 95% confidence dose.
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Table 6-25. Uncertainty for loose-chip TLD beta dose.

Dose (mrad)

Upper 95% confidence dose (mrad), 

1969–1982

1 21

2 22

5 25

10 30

20 40

50 72

100 126

200 239

500 585

1,000 1,166

2,000 2,330

6.8.5.2.2 Panasonic TLD Dosimeter

Table 6-26 lists the uncertainty associated with DOELAP-accredited Panasonic dosimeter dose 

readings. These values were calculated using the TLD uncertainty methodology described in 

Section 2.1.1.3.2 of NIOSH (2007). This method recognizes that the elements of the uncertainty are 

quantified in the dosimetry program documentation available for a DOELAP-accredited program. The 

standard deviation for null readings is from a study performed at RFP (KHC 2001), and the relative 

standard deviation at high readings is the standard deviation of DOELAP performance test results 

(KHC 2001; Stanford 1990). The reasonable worst case value (high-energy photons + neutrons 

mixture from KHC 2001, Table 11) was selected to provide a result that is favorable to claimants.

Table 6-26. Uncertainty for DOELAP-accredited TLD beta dose (upper 95% confidence dose in 

mrem). 

Dose (mrem)

Panasonic dosimeter, 

1983–1989

DOELAP-accredited

Panasonic dosimeter, 

1990–1998

DOELAP-accredited 

Panasonic dosimeter, 

1999–2004

1 1.19 1.19 1.19

2 2.29 2.29 2.28

5 6a 6a 6a

10 11 11 11

20 22 22 22

50 56 56 56

100 112 112 112

200 224 224 223

500 561 561 558

1,000 1,122 1,122 1,116

2,000 2,243 2,243 2,231

a. 1.12 multiplier for any dose greater than 2 mrem.

6.8.6 Skin Contamination

Skin contamination incidents were routinely reported at RFP on a contamination report. Information 

generally indicates the location of the skin contamination and the initial count. The area of the 

contamination might not be available and should be estimated in the manner described in 

Section 2.3.3 of NIOSH (2007).

Depleted uranium is the RFP production material that would result in the greatest skin dose from 

surface contamination. The progeny potentially contained in the material would result in a beta 

exposure to the skin.
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The contamination reports do not indicate the length of time that the contamination was present on 

the skin. An assumption that is favorable to claimants is that the contamination was present for 4 hr. 

This is a reasonable worst case assumption that, for example, the individual received contamination 

at the beginning of the shift, did not take a midmorning break, and discovered the contamination 

during monitoring when leaving the production area at lunch. Once the contamination was 

discovered, initial decontamination would be performed in the production building, which would result 

in removal of most of the contamination. Before 1970, self-monitoring equipment was not readily 

available, and an assumption favorable to claimants of 8 hr is appropriate.

Values in the contamination reports are typically in counts per minute. RFP typically used a GeigerMüller pancake probe to perform uranium surveys. 

DU consists of 99.8% 238U by weight. Table 6-27 lists the other isotopes.

Table 6-27. DU mixtures (DOE 1980).a

Isotope

Ci/g (mix), 

alpha

Ci/g (mix), 

beta

nCi/g 

(mix)

Th-231 N/A 4.90E-09 N/A

Th-234 N/A 3.40E-07 340

U-234 3.70E-08 N/A 37

U-235 4.90E-09 N/A 4.9

U-238 3.40E-07 N/A 340

Total 3.82E-07 3.45E-07 726.8

a. N/A = not applicable.

It is favorable to claimants to assume that the DU is 1 year old. This allows for ingrowth of progeny to 

achieve secular equilibrium. A decay calculation using MicroShield 5.03 (Grove Engineering 1998) 

was performed. Table 6-28 lists the full set of decay isotopes. 

Table 6-28. Activity per gram, 1-year-old DU. 

Nuclide Ci Bq

Ac-227 1.63E-15 6.04E-05

Bi-210a 6.69E-19 2.47E-08

Bi-211 1.28E-15 4.75E-05

Bi-214a 7.00E-17 2.59E-06

Fr-223 2.25E-17 8.34E-07

Pa-231 1.04E-13 3.84E-03

Pa-234a 5.44E-10 2.01E+01

Pa-234m 3.40E-07 1.26E+04

Pb-210a 7.10E-19 2.63E-08

Pb-211 1.28E-15 4.75E-05

Pb-214a 7.00E-17 2.59E-06

Po-210 2.16E-19 7.99E-09

Po-211 3.50E-18 1.30E-07

Po-214 6.99E-17 2.59E-06

Nuclide Ci Bq

Po-215 1.28E-15 4.75E-05

Po-218 7.00E-17 2.59E-06

Ra-223 1.28E-15 4.75E-05

Ra-226 7.21E-17 2.67E-06

Rn-219 1.28E-15 4.75E-05

Rn-222 7.00E-17 2.59E-06

Th-227 1.39E-15 5.15E-05

Th-230 3.33E-13 1.23E-02

Th-231a 4.90E-09 1.81E+02

Th-234a 3.40E-07 1.26E+04

Tl-207 1.28E-15 4.74E-05

U-234 3.70E-08 1.37E+03

U-235 4.90E-09 1.81E+02

U-238 3.40E-07 1.26E+04

a. Significant progeny (included in VARSKIN Mod 3). 

Dose calculation might utilize software such as VARSKIN (recommended in NIOSH 2007) or other 

appropriate means.
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6.9 UNMONITORED INDIVIDUALS

6.9.1 In Production Areas

In the early 1950s only groups expected to receive doses greater than 10% of the radiation protection 

guideline (called the threshold dose at RFP) would receive dosimeters. Attachment C documents the 

RFP external dosimetry coworker study. The data from the coworker study should be used to 

address unmonitored dose for this group of workers. 

6.9.2 Outside Production Areas

After about 1990, many individuals at RFP who did not work in radiological areas were not badged. 

The site radiological protection organization determined that these individuals were unlikely to exceed 

100 mrem of occupational exposure in a CY. 

For individuals who worked outside the radiologically controlled areas, environmental exposure would 

be a better estimate of their exposure [see the latest version of Section 4.0, Environmental Dose, of 

this site profile (ORAUT 2007a)].

6.10 EXTREMITY DOSIMETRY

Extremity dosimeters were used at RFP. Between 1951 and 1970, the site used an ORNL-designed 

film dosimeter similar to that used for the body badge (Baker 2002). The dosimeter was worn on the 

wrist and modified with a brass filter similar to the body badge. Little performance information is 

available on this badge, but it probably performed similarly to the body badge of that period.

In 1971, RFP switched to an in-house-designed wrist dosimeter with four Harshaw chips (Link and 

Pennock 1983; Baker 2002). This badge contained two TLD-600 and two TLD-700 chips that enabled 

neutron and photon dose determination. Uranium workers received an OW (thin Mylar) version.

In 1991, RFP switched to a Panasonic model UD-813AS11 (custom design) dosimeter in a plastic 

wrist holder (KHC 2000b; Baker 2002). This dosimeter contains two 6

Li-borate elements and two 

7

Li-borate elements that enable neutron dose measurement. Two of the elements are under a thin 

OW for beta and low-energy photon dose measurements. The dosimeter, which has undergone 

DOELAP performance testing, is documented in KHC (2000b).

RFP did not use finger rings; however, wrist badges were used to assess the dose to the forearm. 

Hand dose was assessed using the dose measured by the wrist badge and the application of a handto-wrist ratio. Falk (1976) documents hand-to-wrist ratios of 1.5 for Buildings 771 and 559, and 2.5 for 

all other buildings. Section 05.04 of KHC (2000b) indicates a ratio of 3 was implemented in 

approximately 1992 as a conservative estimate based on the results of several studies.

Many RFP workers did not receive extremity (wrist) dosimeters. In such cases, the wrist (forearm) 

dose was assigned as the measured skin (shallow) dose, and the hand dose was assigned the same 

value. If an extremity dosimeter was worn and the value was less than the skin dose measured by 

the body badge, the assumption was made that the extremity dosimeter was not worn and the skin 

dose was assigned as the wrist dose. If the extremity dosimeter did measure a dose greater than the 

body badge, the extremity measurement was assigned to the wrist and a hand-to-wrist ratio was used 

to estimate the dose to the hand. Several studies over the years determined the hand-to-wrist ratio 

(Falk 1976; KHC 2000b). 

Additional information on these dosimeters will be required for dose reconstruction for shallow dose to 

the extremity, if necessary.
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6.11 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
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REFERENCES

Alvarez, J., L. Faust, J. Fix, D. Habib, G. Kerr, J. Langsted, E. Lesses, J. Martin, J. Mohrbacher, 

N. Rohrig, and T. Widner, 2003, MED/AEC/DOE External Dosimetry Technology Technical 

Basis Document, Rev. 0, Oak Ridge Associated Universities, Oak Ridge, Tennessee, 

August 21. [SRDB Ref ID: 4618]

Author unknown, 1993, Basis and Effects of the Inclusion of Minimum Detectable Dose Calculations 

in the Latest Revision of the Personnel Dose Algorithm on Personnel Doses at Rocky Flats 

Plant, November 11. [SRDB Ref ID: 4626]

Author unknown, undated, Algorithm for Panasonic 802 and 809 Dosimeters for the Rocky Flats 

Dosimetry Badge. [SRDB Ref ID: 4627]

Baker, S., 2002, “Historical Timeline of External Dosimetry Equipment Used at Rocky Flats Since 

1952,” memorandum to External Dosimetry File, Kaiser Hill Company, Golden, Colorado, 

February 12. [SRDB Ref ID: 1251]

Ballenger, R. J., 1992, “Annual Review of Operational Safety Analysis (OSA) 779.114 Gamma 

Irradiation Cell, Building 779,” memorandum to S. L. Nortnik, EG&G Rocky Flats, Rocky Flats 

Plant, Golden Colorado, September 17. [SRDB Ref ID: 165245, p. 8]

Barker, J. J., editor, 1969, Californium-252, Proceedings of a Symposium Sponsored by The New 

York Metropolitan Section of the American Nuclear Society, New York City, October 22, 1968,

CONF-681032, U.S. Atomic Energy Commission, Division of Technical Information, January. 

[SRDB Ref ID: 11502]

Brackenbush, L. W., F. M. Cummings, G. W. R. Endres, J. E. Tanner, and M. P. Moeller, 1989, 

Technical Assistance to Rocky Flats Plant, Neutron and Photon Dose Measurements, Battelle 

Memorial Institute, Pacific Northwest Laboratory, Richland, Washington, April. [SRDB Ref ID: 

4642]

DOE (U.S. Department of Energy), 1980, Final Environmental Impact Statement (Final Statement to 

ERDA 1545-D), Rocky Flats Plant Site, Golden, Jefferson County, Colorado, DOE/EIS-0064, 

U.S. Department of Energy, Washington, D.C., April. [SRDB Ref ID: 11505, 11507, 11508]

DOE (U.S. Department of Energy), 1992, U.S. Department of Energy Radiological Control Manual, 

DOE/EH-0256T, DOE N 5480.6, Office of Assistant Secretary for Environment, Safety and 

Health, Washington, D.C., June. [SRDB Ref ID: 11515]

DOE (U.S. Department of Energy), 2001, Guide of Good Practices for Occupational Radiological 

Protection in Uranium Facilities, DOE-STD-1136-2000, Change Notice 3, Washington, D.C., 

December. [SRDB Ref ID: 4617]

DOE (U.S. Department of Energy), 2003, Guide of Good Practices for Occupational Radiological 

Protection in Plutonium Facilities, DOE-STD-1128-98, Reaffirmation with Errata, Washington, 

D.C., May. [SRDB Ref ID: 15919]

Falk, R. B., 1971, A Personnel Neutron Dosimeter Using Lithium Fluoride Thermoluminescent 

Dosimeters, RFP-1581, The Dow Chemical Company, Rocky Flats Plant, Golden, Colorado, 

January 20. [SRDB Ref ID: 1272]
Document No. ORAUT-TKBS-0011-6 Revision No. 03 Effective Date: 01/14/2019 Page 65 of 109

 

Falk, R. B., 1975, “PuF4 Update,” handwritten notes, Rockwell International, Rocky Flats Plant, 

Golden, Colorado, May 22. [SRDB Ref ID: 4636]

Falk, R. B., 1976, TLD Calculations, Rockwell International, Rocky Flats Plant, Golden, Colorado, 

December 8. [SRDB Ref ID: 1268, p. 2] 

Falk, R. B., undated, “Conversion of β/γ Film Sector Doses to Penetrating Photon Dose,” Rocky Flats 

Plant, Golden, Colorado. [SRDB Ref ID: 4635]

Falk, R. B., J. M. Aldrich, J. Follmer, N. M. Daugherty, D. E. Hilmas, and P. L. Chapman, 2005, 

Technical Basis Document for the Neutron Dose Reconstruction Project, Neutron Dose 

Reconstruction Protocol, ORISE 05-0199, Oak Ridge Institute of Science and Education, 

Oak Ridge, Tennessee, February 7. [SRDB Ref ID: 17126]

ICRP (International Commission on Radiological Protection), 1991, 1990 Recommendations of the 

International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England. [SRDB Ref ID: 11517]

ICRP (International Commission on Radiological Protection), 1996, Conversion Coefficients for Use in 

Radiological Protection Against External Radiation, Publication 74, Pergamon Press, Oxford, 

England. [SRDB Ref ID: 7979]

Jens, J. P., 1990, “Administrative Changes in Radiation Dosimeter Management - Date Correction,” 

memorandum to All EG&G Employees et al., EG&G Rocky Flats, Rocky Flats Plant, Golden 

Colorado, December 4. [SRDB Ref ID: 24796]

Kazanjian, A. R., and A. K. Brown, 1969, Radiation Chemistry of Materials Used in Plutonium 

Processing, RFP-1376, Dow Chemical Company, Rocky Flats Division, Golden, Colorado, 

September 4. [SRDB Ref ID: 160151]

KHC (Kaiser-Hill Company), 2000a, Report on the Flow of Recycled Uranium at the Department of 

Energy's Rocky Flats Plant, 1953 - 1993, Rocky Flats Environmental Technology Site, Golden, 

Colorado, June 30. [SRDB Ref ID: 16227, p. 5] 

KHC (Kaiser-Hill Company), 2000b, Rocky Flats Environmental Technology Site External Dosimetry 

Technical Basis Document - Extremity Dosimetry, ED-TBM-05, Rev. 02, Rocky Flats 

Environmental Technology Site, Golden, Colorado, September 25. [SRDB Ref ID: 8797, 

p. 59]

KHC (Kaiser-Hill Company), 2001, External Dosimetry Technical Basis Document - Whole Body 

Dosimetry, ED-TBM-04, Rev. 2, Rocky Flats Environmental Technology Site, Golden, 

Colorado, September 25. [SRDB Ref ID: 8797, p. 34]

Lalos, G., editor, 1989, Film Badge Dosimetry in Atmospheric Nuclear Tests, National Research 

Council, National Academy Press, Washington, D.C. [SRDB Ref ID: 1905]

Lombardi, E. F., and R. I. Donovan, 1971, An Evaluation of Fiberglas for Use in Cushioning Plutonium 

Parts, RFP-1725, Dow Chemical Company, Rocky Flats Division, Golden, Colorado, 

November 24. [SRDB Ref ID: 160152]

Link, R. A., and J. R. Pennock, 1983, External Thermoluminescence Dosimetry (TLD), HS&EL OP-39, 

Rockwell International, Rocky Flats Plant, Golden, Colorado, April 29. [SRDB Ref ID: 4633]
Document No. ORAUT-TKBS-0011-6 Revision No. 03 Effective Date: 01/14/2019 Page 66 of 109

 

Mann, J. R., 1964, “Handling and Processing Dosimeter Film at Rocky Flats Division,” memorandum 

to J. B. Owen, The Dow Chemical Company, Rocky Flats Plant, Golden, Colorado, August 3. 

[SRDB Ref ID: 4628]

Mann, J. R., 1967, Film Badge Dosimetry Calibration Procedure, memorandum to Radiation 

Dosimetry Technicians, The Dow Chemical Company, Rocky Flats Plant, Golden, Colorado, 

January 18. [SRDB Ref ID: 4640]

Mann, J. R., and M. R. Boss, 1963, Neutron Dosimetry Problems in a Plutonium Processing Plant, 

RFP-305, The Dow Chemical Company, Rocky Flats Division, Denver, Colorado. [SRDB Ref 

ID: 4629]

NBS (National Bureau of Standards), 1957, Protection Against Neutron Radiation up to 30 Million 

Electron Volts, Handbook 63, U.S. Department of Commerce, Washington, D.C., November 

22. [SRDB Ref ID: 10025]

NCRP (National Council on Radiation Protection and Measurements), 1971, Protection Against 

Neutron Radiation, NCRP Report 38, Bethesda, Maryland, January 4. [SRDB Ref ID: 11520]

NIOSH (National Institute for Occupational Safety and Health), 2007, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 3, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, November 21. [SRDB Ref ID: 38864]

NIOSH (National Institute for Occupational Safety and Health), 2010, Radiation Exposures Covered 

for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness

Compensation Program Act, DCAS-IG-003, Rev. 1, Division of Compensation Analysis and 

Support, Cincinnati, Ohio, October 5. [SRDB Ref ID: 88929]

NIOSH (National Institute for Occupational Safety and Health), 2011, Best Estimate External Dose 

Reconstruction for Glovebox Workers, DCAS-TIB-0010, Rev. 04, Cincinnati, Ohio, 

November 28. [SRDB Ref ID: 104815]

Nogues, H., 1999, “Site’s largest radiation source cleared out, B779 demolition can get under way in 

October,” Rocky Flats Envision, volume 5, number 19. [SRDB Ref ID: 165233]

ORAUT (Oak Ridge Associated Universities Team), 2005a, Savannah River Site, ORAUT-TKBS0003, Rev. 03, Oak Ridge, Tennessee, April 5. [SRDB Ref ID: 20176]

ORAUT (Oak Ridge Associated Universities Team), 2005b, Interpretation of Dosimetry Data for 

Assignment of Shallow Dose, ORAUT-OTIB-0017, Rev. 01, Oak Ridge, Tennessee, 

October 11. [SRDB Ref ID: 19434]

ORAUT (Oak Ridge Associated Universities Team), 2006a, A Standard Methodology for 

Overestimating External Doses Measured with Thermoluminescent Dosimeters, ORAUTOTIB-0008, Rev. 01, Oak Ridge, Tennessee, May 12. [SRDB Ref ID: 29953]

ORAUT (Oak Ridge Associated Universities Team), 2006b, Occupational Onsite Ambient Dose 

Reconstruction for DOE Sites, ORAUT-PROC-0060, Rev. 01, Oak Ridge, Tennessee, 

June 28. [SRDB Ref ID: 29986]

ORAUT (Oak Ridge Associated Universities Team), 2007a, Rocky Flats Plant – Occupational 

Environmental Dose, ORAUT-TKBS-0011-4, Rev. 02, Oak Ridge, Tennessee, April 23. 

[SRDB Ref ID: 31377]
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ORAUT (Oak Ridge Associated Universities Team), 2007b, Oak Ridge National Laboratory – 

Occupational External Dose, ORAUT-TKBS-0012-6, Rev. 01, Oak Ridge, Tennessee, 

September 10. [SRDB Ref ID: 34945]

ORAUT (Oak Ridge Associated Universities Team), 2011, Use of Coworker Dosimetry Data for 

External Dose Assignment, ORAUT-OTIB-0020, Rev. 03, Oak Ridge Tennessee, 

November 14. [SRDB Ref ID: 104029]

ORAUT (Oak Ridge Associated Universities Team), 2013, Los Alamos National Laboratory – 

Occupational External Dose, ORAUT-TKBS-0010-6, Rev. 03, Oak Ridge, Tennessee, 

March 21. [SRDB Ref ID: 123074]

ORAUT (Oak Ridge Associated Universities Team), 2014, Parameters to Consider When Processing 

Claims for Construction Trade Workers, ORAUT-OTIB-0052, Rev. 02, Oak Ridge, Tennessee, 

July 24. [SRDB Ref ID: 133862] 

ORAUT (Oak Ridge Associated Universities Team), 2018a, Documented Communication With [name 

redacted] on ORAUT-TKBS-0011-6 A&A Validation, Oak Ridge, Tennessee, September 19. 

[SRDB Ref ID: 174141] 

ORAUT (Oak Ridge Associated Universities Team), 2018b, External Dose Reconstruction, ORAUTOTIB-0088, Rev. 00, Oak Ridge, Tennessee, September 25. [SRDB Ref ID: 174087] 

Passmore, C. N., 1992, RFP Lead Apron Field Study (TLD Inside vs. Outside Lead Apron), Rockwell 

International, Rocky Flats Plant Site, Golden, Colorado, March 21. [SRDB Ref ID: 22289]

Patel, J. S., 1992, Radioactive Materials Associated with Rocky Flats Plant, Monthly Information 

Exchange Meeting, Rocky Flats Plant, Golden, Colorado, October 27. [SRDB Ref ID: 

165239]

Putzier, E. A., 1982, The Past 30 Years at Rocky Flats Plant, Rockwell International, Rocky Flats 

Plant, Golden, Colorado, November. [SRDB Ref ID: 4632]

RFETS (Rocky Flats Environmental Technology Site), 1993, Registry of Radioactive Sealed Sources 

and Devices, Safety Evaluation of Device, NR-220-D-107-S, Rocky Flats Plant, Golden, 

Colorado, June 29. [SRDB Ref ID: 165234]

RFETS (Rocky Flats Environmental Technology Site), 1993–2002, Cobalt-60 Freedom of Information 

Act Request, Rocky Flats Plant, Golden, Colorado. [SRDB Ref ID: 148917]

RFETS (Rocky Flats Environmental Technology Site), 1998a, Post Job Review Checklist & 

Instructions, Cobalt 60 Source Removal, Project Number T0100734, Rocky Flats Plant, 

Golden, Colorado, September 15. [SRDB Ref ID: 148917, pp. 100–101]

RFETS (Rocky Flats Environmental Technology Site), 1998b, ALARA Job Review Narrative, Rev. 0, 

Log No.: 99-779-012, Rocky Flats Plant, Golden, Colorado, March. [SRDB Ref ID: 148917, 

p. 84]

RFETS (Rocky Flats Environmental Technology Site), 1998c, Job Specific Contamination Survey 

Source Leak Test, Building 779, Room 218, Rocky Flats Plant, Golden, Colorado, August 13. 

[SRDB Ref ID: 165242]
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RFETS (Rocky Flats Environmental Technology Site), 1998d, Job Specific Survey, Building 779, 

Room 218, Rocky Flats Plant, Golden, Colorado, February 9. [SRDB Ref ID: 165244]

RFETS (Rocky Flats Environmental Technology Site), 1999a, Activity Screening Form: Remove 

Cobalt 60 Source from Room 218, Rocky Flats Plant, Golden, Colorado, April. [SRDB Ref ID: 

148917, p. 14]

RFETS (Rocky Flats Environmental Technology Site), 1999b, Engineering Order: Cobalt 60 Source 

Removal, Project Number T0100734, Rocky Flats Plant, Golden, Colorado, May 24. [SRDB 

Ref ID: 148917, pp. 5–275]

RFETS (Rocky Flats Environmental Technology Site), 1999c, Radioactive Source Report, Reg. 

No. 105, Rocky Flats Plant, Golden, Colorado, June 17. [SRDB Ref ID: 165236]

RFETS (Rocky Flats Environmental Technology Site), 1999d, Job Specific Contamination Survey 

Source Inventory and Audit, Buildings 779 and 706, Rooms 218 and 124, Rocky Flats Plant, 

Golden, Colorado, January 27. [SRDB Ref ID: 165243]

RFETS (Rocky Flats Environmental Technology Site), undated a, Work Package Continuation: 

Cobalt 60 Source Removal Building 779, Project Number T0100734, Rocky Flats Plant, 

Golden, Colorado. [SRDB Ref ID: 148917, p. 35]

RFETS (Rocky Flats Environmental Technology Site), undated b, Design Work Special Instructions: 

Cobalt 60 Source Removal, Project Number T0100734, Rocky Flats Plant, Golden, Colorado. 

[SRDB Ref ID: 148917, p. 34]

RFETS (Rocky Flats Environmental Technology Site), undated c, Operational Safety Analysis, 

Hazard/Hazard Control, Rocky Flats Plant, Golden, Colorado. [SRDB Ref ID: 165245, pp. 4–

6]

Rockwell International, 1985, Radiation Exposure to Rockwell Employees at the Rocky Flats Plant, 

Annual Report -- Year 1984, Rocky Flats Plant, Golden, Colorado. [SRDB Ref ID: 7958]

Rockwell (Rockwell International), 1987, Operational Health Physics’ Radioactive Source Audit 

Report, Rocky Flats Plant, Golden, Colorado, February. [SRDB Ref ID: 165240]

Ruttenber, A. J., M. Schonbeck, S. Brown, T. Wells, D. McClure, J. McCrea, D. Popken, J. Martyny, 

2003, Report of Epidemiologic Analyses Performed for Rocky Flats Production Workers 

Employed Between 1952-1989, University of Colorado Health Sciences Center, and Colorado 

Department of Public Health and Environment, March 3. [SRDB Ref ID: 8060]

Savitz, K., undated, Explanation of the Electronic Data, Rocky Flats Plant, Golden, Colorado. [SRDB 

Ref ID: 24757]

Shepherd, J., 1999, “GC 220 Tie Down Procedures,” letter to T. Westly (Rocky Mountain Remediation 

Services), J. L. Shepherd & Associates, San Fernando, California, May 21. [SRDB Ref ID: 

148917, p. 150]

Stanford, N., 1990, Whole Body TLD Dose Algorithm for Panasonic 802/809 TLD Pair, Rockwell 

International, Rocky Flats Plant, Golden Colorado, September 30. [SRDB Ref ID: 4641]
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Wilson, R. H., J. J. Fix, W. V. Baumgartner, and L. L. Nichols, 1990, Description and Evaluation of the 

Hanford Personnel Dosimeter Program from 1944 through 1989, PNL-7447, Battelle Memorial 

Institute, Pacific Northwest Laboratory, Richland, Washington, September. [SRDB Ref ID: 

4793]
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GLOSSARY

beta dose

Designation (i.e., beta) on some records for external dose from beta and less-energetic X-ray 

and gamma radiation, often for shallow dose or dose to the lens of the eye.

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.

deep absorbed dose

Absorbed dose in units of rem or sievert at a depth of 1 centimeter (1,000 milligrams per 

square centimeter). See dose. 

deep dose equivalent (DDE)

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

per square centimeter). See dose. 

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, rad, rep, or grays. 

dose equivalent (H) 

n 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. See film dosimeter, neutron film dosimeter, and thermoluminescent 

dosimeter. 

dosimetry

Measurement and calculation of internal and external radiation doses.

dosimetry system

System for assessment of received radiation dose. This includes the fabrication, assignment, 

and processing of external dosimeters, and/or the collection and analysis of bioassay samples, 

and the interpretation and documentation of the results.

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; see acute exposure and chronic 

exposure. (2) Measure of the ionization produced by X- and gamma-ray photons in air in units 

of roentgens.
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extremities

Portion of the arm from and including the elbow through the fingertips and the portion of the 

leg from and including the knee and patella through the toes.

field calibration

Dosimeter calibration based on radiation types, intensities, and energies in the work 

environment.

film

In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. See film dosimeter. 

film density

See optical density. 

film dosimeter

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

fission

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. 

fissionable

Capable of undergoing fission by capturing neutrons, including fast neutrons. Uranium-238 is 

fissionable. Fissionable indicates both spontaneous and induced fission.

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. 

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 

beta radiation and gamma 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.
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neutron

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

neutron film dosimeter

Film dosimeter with a nuclear track emulsion, type A, film packet.

nuclear track emulsion, type A (NTA)

Film made by the Eastman Kodak Company that is sensitive to fast neutrons. The developed 

image has tracks caused by neutrons that become visible under oil immersion with about 

1,000-power magnification. The number of tracks in a given area is a measure of the dose 

from that radiation.

open window (OW)

Area of a film dosimeter that has little to no radiation shielding (e.g., only a holder and visible 

light protection). The open window measures nonpenetrating as well as penetrating dose, 

which minimizes the potential for beta radiation to contribute to the interpreted penetrating 

dose. See film dosimeter. 

operating area

Usually refers to a major operational work area at a site.

optical density

Measure of the degree of opacity of photographic or radiographic film defined as OD = 

log10 (I0/I), the base-10 logarithm of the ratio of the reference light intensity I0 (without film) to 

the transmitted light intensity (through the film) Also called film density and density reading.

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. In 1993 the International Commission on 

Radiological Measurement and Units recommended Hp(d) as the dose quantity for radiological 

protection.

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 X-ray

Electromagnetic radiation of energies between 10 keV and 100 keV whose source can be an 

X-ray machine or radioisotope.

pit

Package at the center of an implosion weapon that contains the machined fissile material that 

begins the fission chain reaction. Also called primary.

quality factor (Q, 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.
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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. 

radionuclide

Radioactive nuclide.

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.

roentgen (R, sometimes 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 degrees Celsius 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 absorbed dose (Ds) 

Absorbed dose at a depth of 0.07 millimeters (7 milligrams per square centimeter) in a material 

of specified geometry and composition.

shallow dose equivalent [SDE, Hs, Hp(0.07)] 

Dose equivalent in units of rem or sievert at a depth of 0.07 millimeters (7 milligrams per 

square centimeter) in tissue equal to the sum of the penetrating and nonpenetrating doses.

shielding

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

silver shield

The 1-mm-thick shields covering the film packet in early film dosimeters. 

skin dose

See shallow dose equivalent. 

thermoluminescence

Property that causes a material to emit light as a result of heat.

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.
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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.

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 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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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

LIST OF FIGURES 

FIGURES TITLE PAGE

A-1 Occupational Dose Report reviewed June 4, 2002................................................................. 76

A-2 Occupational Dose Report reviewed June 4, 2002................................................................. 77

A-3 Dosimetry History by Individual query report dated March 10, 2003....................................... 78

A-4 Dosimetry History by Individual query report dated March 10, 2003....................................... 79

A-5 Dosimetry History by Individual query report dated March 10, 2003....................................... 80

A-6 Health Physics Yearly External Exposure Activity Run, 1953 to 1958 .................................... 81

A-7 Health Physics Yearly External Exposure Run, 1959 to 1963 ................................................ 82

A-8 Health Physics Yearly External Exposure Activity Run, 1964................................................. 83

A-9 External Dosimetry Detail, 1981 to 1983 ................................................................................ 84

A-10 Health Physics External Radiation Exposure Report, 1967 .................................................... 85

A-11 Radiation Dosimetry Individual Lifetime Report dated June 4, 2002....................................... 86

A-12 Radiation Health Records System – TLD Data....................................................................... 87

A-13 Radiation Dosimetry Detail Report, Termination Report dated September 17, 1996 .............. 88

A-14 Radiation Dosimetry Detail Report, Termination Report dated September 17, 1996 .............. 89

A-15 Occupational Radiation Exposure Information, 1988.............................................................. 90
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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-1. Occupational Dose Report reviewed June 4, 2002. 
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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-2. Occupational Dose Report reviewed June 4, 2002. 
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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-3. Dosimetry History by Individual query report dated March 10, 2003. 
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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-4. Dosimetry History by Individual query report dated March 10, 2003. 
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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-5. Dosimetry History by Individual query report dated March 10, 2003. 
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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-6. Health Physics Yearly External Exposure Activity Run, 1953 to 1958. 
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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-7. Health Physics Yearly External Exposure Run, 1959 to 1963. 
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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-8. Health Physics Yearly External Exposure Activity Run, 1964. 
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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-9. External Dosimetry Detail, 1981 to 1983. 
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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-10. Health Physics External Radiation Exposure Report, 1967. 
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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-11. Radiation Dosimetry Individual Lifetime Report dated June 4, 2002. 
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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-12. Radiation Health Records System – TLD Data. 
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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-13. Radiation Dosimetry Detail Report, Termination Report dated September 17, 1996. 
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ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-14. Radiation Dosimetry Detail Report, Termination Report dated September 17, 1996. 
Document No. ORAUT-TKBS-0011-6 Revision No. 03 Effective Date: 01/14/2019 Page 90 of 109

ATTACHMENT A 

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS (continued)

Figure A-15. Occupational Radiation Exposure Information, 1988. 
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ATTACHMENT B 

MAJOR JOB CATEGORIES

Chemical Operators

Primary job duties included highly enriched uranium (HEU) (Building 881) and plutonium 

(Building 771/371) metal reprocessing using dissolution, fluorination, calcine, and other wet chemistry 

methods to purify metal in preparation for foundry casting operations. Molten salt processing 

(Building 776) was an exceptionally high neutron process. Other typical job duties included waste 

treatment (Building 774/374) for waste solutions generated across RFP.

Metallurgical Operators

Primary job duties included casting (Building 881), rolling and pressing HEU (Building 883), plutonium 

(Building 776/ 707), and DU (B444/447 and 883). Exposures tended to be less than those to 

Chemical Operators. Machinists, Assemblers, Material Analysts, and Welders had similar exposures. 

Nondestructive Testing Technicians had similar, but probably lower, exposures because work was 

often done on completed pits that inherently shielded fissile materials. Experimental Operators had 

similar, but probably higher, exposures because they often worked with prototype systems or 

processes that lacked shielding and other radiological controls as the regular production processes.

Maintenance Workers

Typical trades (i.e., machinists, pipefitters, welders, carpenters, painters, electricians) had varied 

exposures because they often did more intrusive work on contaminated systems than production 

personnel. Examples of intrusive work include repairing leaks on process lines (pipefitters), refractory 

replacement in casting and heat treat furnaces (carpenters), repair of mechanical systems 

(machinists) and repair of instruments and controllers inside gloveboxes and other systems 

(electricians), painting over contamination (painters).

Support Personnel

This category includes Clerk Packers, Metrology Technicians, Janitors, and Handymen, who worked 

in process areas but did little or no hands-on work with radioactive materials. Exposures would be 

incidental to working in rooms with process equipment (metallurgical and chemical operations).

Analytical Laboratory Technicians

These individuals worked primarily in Building 559 (plutonium samples) or Building 881 (HEU/DU 

samples) and probably had lower exposures than operators who performed hands-on work with 

significantly higher material quantities.

Site Support Personnel

Stationary Operating Engineers (SOEs, also known as Boiler Vent Operators), Security Guards, Shift 

Managers, and Configuration Control Authority personnel performed little if any hands-on work, but 

had routine access to process areas. SOEs monitored exhaust systems, waste tanks, and process 

waste lines. Exposures would be incidental to working in rooms with process equipment 

(metallurgical and chemical operations).

Radiation Control Technicians

Radiation Control Technicians probably had exposures from supporting production chemical and 

metallurgical processes. Some exposures probably occurred during decontamination activities, 

surveys of contaminated areas, upset conditions. They generally performed no hands-on work, but 

typically worked side-by-side with production operators.
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ATTACHMENT B 

MAJOR JOB CATEGORIES (continued)

D&D Workers

D&D work includes draining actinide systems, decontamination, size reduction and removal of 

contaminated equipment, gloveboxes, piping, ductwork, exhaust systems, waste packaging of 

removed equipment, low-level, and transuranic wastes. Work is often in high (>2,000 dpm 

removable) airborne contamination areas with Derived Air Concentration levels from >0.1 to 106. 

Personal protection equipment includes Air Purifying Respirator, Powered Air Purifying Respirator, or 

PremAir supplied air. There were some high exposures due to direct work with highly radioactive 

equipment and contamination events.
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ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT

TABLE OF CONTENTS

SECTION TITLE PAGE

C.1 General Approach.................................................................................................................. 94

C.2 Applications and Limitations................................................................................................... 94

C.3 Coworker Data Development ................................................................................................. 94

C.4 Adjustment for Missed Dose .................................................................................................. 95

C.4.1 Plutonium Workers (Before 1971)............................................................................... 95

C.4.2 Plutonium (After 1970) and Uranium (all years) Workers............................................ 95

C.5 Special Considerations .......................................................................................................... 96

C.6 Coworker Annual Dose Summaries ....................................................................................... 97

C.7 Plutonium Workers................................................................................................................. 97

C.7.1 1952 to 1970 .............................................................................................................. 97

C.7.2 1971 to 2005 .............................................................................................................. 98

C.8 Uranium Workers ................................................................................................................... 99

C.9 Coworker Dose Tables......................................................................................................... 101

LIST OF TABLES

TABLE TITLE PAGE

C-1 Missed external doses ........................................................................................................... 96

C-2 Lognormal neutron-to-photon ratio values, 1970 to 1976 ....................................................... 97

C-3 Annual external coworker doses for plutonium workers, 1952 to 1970 ................................. 101

C-4 Annual external coworker doses for plutonium workers, 1971 to 2005 ................................. 103

C-5 Annual external coworker doses for uranium workers, 1952 to 2005.................................... 104

C-6 Annual external coworker doses for plutonium workers, 1952 to 1970, modified in 

accordance with ORAUT-OTIB-0052 ................................................................................... 105

C-7 Annual external coworker doses for plutonium workers, 1971 to 2005, modified in 

accordance with ORAUT-OTIB-0052 ................................................................................... 107

C-8 Annual external coworker doses for plutonium workers, 1952 to 2005, modified in 

accordance with ORAUT-OTIB-0052 ................................................................................... 108
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ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

C.1 GENERAL APPROACH

As described in ORAUT-OTIB-0020, Use of Coworker Dosimetry Data for External Dose Assignment, 

the general approach to the development of coworker data for cases without external monitoring data 

is to assign either 50th- or 95th-percentile doses with the intent that the assigned doses represent, but 

do not underestimate, the doses that would have been assigned if the worker had been monitored

(ORAUT 2011). 

C.2 APPLICATIONS AND LIMITATIONS

Some RFP workers could have worked at one or more other major sites in the DOE complex during 

their employment histories. Therefore, the data in this attachment must be used with caution to 

ensure that, for likely noncompensable cases, unmonitored external doses from multiple site 

employments have been overestimated. This typically requires the availability of the recorded doses 

or external coworker dosimetry data for all relevant sites.

The data in this attachment address penetrating radiation from gamma and neutron radiation and 

nonpenetrating radiation from electron and/or low-energy photon radiation. External onsite ambient 

dose should be applied as specified in the latest revision of ORAUT-PROC-0060, Occupational 

Onsite Ambient Dose Reconstruction for DOE Sites (ORAUT 2006b).

Neutron coworker doses before 1967 are not covered in this attachment. In conjunction with SEC 

Petition Evaluation Report, Petition SEC-00030, neutron coworker doses cannot be created that 

would ensure that a bounding exposure is assigned. Therefore, neutron coworker doses should not 

be assigned before 1967 (NIOSH 2006). 

C.3 COWORKER DATA DEVELOPMENT

Dosimetry data for monitored RFP workers in the HIS20 database and the NDRP database were 

selected for this evaluation. HIS20 was the last system used at the RFP for the retention of 

occupational radiation exposure data. The NDRP database is a compilation of the dosimetry data 

reevaluated by the project and includes gamma and neutron dosimetry data from 1952 to 1970. The 

information in these systems contains data that have been transferred from previous electronic 

systems and hard-copy health physics files (Falk et al. 2005). 

For plutonium workers, the NDRP reported building-specific cycle data between 1952 and 1970. 

These were converted to daily dose rates based on the length of the exchange period. The daily dose 

rates were then prorated to represent a full year of monitored employment. After 1970, HIS20 annual 

data were prorated to account for partial years of employment based on an analysis of the length of 

monitored employment associated with the data. The data were prorated so coworker doses that 

represent a full year of monitored employment could be derived; this permits the dose reconstructor to 

assign appropriate doses based on specific employment dates and job descriptions.

For uranium workers, HIS20 annual data from 1952 to 2005 were prorated to account for partial years 

of employment based on an analysis of the length of monitored employment associated with the data. 

The data were prorated so coworker doses that represent a full year of monitored employment could 

be derived; this permits the dose reconstructor to assign appropriate doses based on specific 

employment dates and job descriptions. 
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ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

The validity of the data that were used for coworker dose development was confirmed by selecting a 

sampling of beta-gamma film badge worksheets (handwritten records) and comparing them to data for 

penetrating radiation listed in the HIS20 database. 

Each beta-gamma worksheet contains film badge results for numerous workers for a given building 

and quarter. For each worker-year, four sheets of quarterly data are combined to comprise the 

annual beta-gamma dose record. Thirty such worker-years were examined and compared to data for 

the same worker-year. 

Of the 30 worker-years compiled (which represent data for 30 individuals), 22 (73%) were complete in 

that all quarterly data were found and the total annual dose was in agreement. For 5 worker-years 

(17%), one-quarter of the beta-gamma worksheet data was not found, but the annual total calculated 

without that missing quarter agreed with the HIS20 database. For 3 worker-years (10%), some 

quarterly beta-gamma worksheet data were not found or were blank and the annual totals did not 

agree with the HIS20 database. In all three of those instances, the HIS20 database annual values 

were higher than the data from the beta-gamma worksheets.

C.4 ADJUSTMENT FOR MISSED DOSE

According to OCAS-IG-001, missed doses are assigned for reported zero readings for each 

monitoring cycle to account for the possibility that doses were received but either not recorded by the 

dosimeter or not reported by the site (NIOSH 2007). In addition, reported dose values less than onehalf the applicable minimum detection limits are assigned as missed dose. Annual maximum potential 

missed doses are calculated by multiplying the number of zero or unrecorded badge readings by the 

reported dosimeter LOD and summing the results. These values are used as the 95th percentile of a 

lognormal distribution to calculate the POC, which is determined by DOL. Therefore, Parameter 1 

input in IREP is equal to the calculated maximum annual missed doses multiplied by 0.5, and the 

Parameter 2 input is equal to 1.52. These values represent the GM and GSD, respectively, for each 

year of analysis.

C.4.1 Plutonium Workers (Before 1971)

For pre-1971 plutonium workers, dose is based on a daily dose rate assigned continuously for every 

day of the year. Therefore, missed dose is not applicable when a calculated dose rate could be 

determined. If the calculated dose rate was determined to be zero, missed dose was assigned for the 

entire year. The average exchange rate was determined based on recorded cycle exchange rates for 

the given building and year. This exchange rate was then used to estimate the expected number of 

badge exchanges for a year.

C.4.2 Plutonium (After 1970) and Uranium (all years) Workers

The assignment of maximum potential missed doses for monitored workers is particularly significant 

for 1954 to 1962, when RFP workers could have been monitored weekly. Table C-1 lists the 

maximum annual missed dose by monitoring period.
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ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Table C-1. Missed external doses (rem).

Monitoring 

period

Penetrating 

LOD

Nonpenetrating 

LODa

Exchange 

frequency

Maximum potential 

annual missed 

penetrating dose

Maximum potential 

annual missed 

nonpenetrating dose

1952–1953 0.04 0.05 Semimonthlyb 0.960 1.200

1954–1962 0.04 0.05 Weeklyb 2.080 2.600

1963–1964 0.04 0.05 Semimonthlyb 0.960 1.200

1965–1966 0.04 0.05 Monthlyb 0.480 0.600

1967–1968 0.04 0.05 Semimonthlyb 0.960 1.200

1969–1976 0.02 0.03 Semimonthlyb 0.480 0.720

1977–1992 0.02 0.03 Monthlyc 0.240 0.360

1993–2005 0.01 0.02 Monthlyc 0.120 0.240

a. Based on analysis of nonpenetrating LODs for other DOE sites in ORAUT-OTIB-0017 (ORAUT 2005b). Specific RFP 

data for nonpenetrating LOD are not available at this time.

b. Based on maximum potential exchange frequency.

c. The exchange frequency was not defined. It is based on a review of claim data evaluated under the EEOICPA 

Subtitle B program.

C.5 SPECIAL CONSIDERATIONS

Certain aspects of the external dosimetry practices at RFP were considered in the analysis of the site 

data. These include:

• Conservatively determined default dosimeter exchange frequencies were used. Not all RFP 

employees would have had dosimetry exchanged at these frequencies

• During the process of prorating HIS20 doses to account for partial years of employment, it was 

discovered that artificially short or long wear periods had been entered in the HIS20 database. 

These artificial periods were entered because only one date (usually the end date) was 

available in the electronic data from previous database systems. To avoid skewing the overall 

dataset with artificially high or low prorated dose, data with a wear period less than or equal to 

0.1 yr or greater than 1.25 year were excluded from the analysis described below.

• Inclusion of NDRP data in the HIS20 database led to the development of tables of data for 

penetrating and nonpenetrating dose that include and exclude the NDRP values. To 

determine neutron dose for the period from 1952 to 1969, neutron-to-photon ratios defined in 

the NDRP study (Falk et al. 2005) were used. Table C-2 defines the neutron-to-photon ratios 

for 1970 to 1976. 

Neutron-to-photon ratio values for 1970 to 1976 were developed by analyzing records that contain 

dosimetry reports to supervisors. Data were available for workers on a semimonthly dosimeter 

exchange cycle for 1971, 1972, 1974, 1975, and 1976. Data for workers on a monthly exchange 

cycle were available for 1972 and 1976. Neutron-to-photon ratio values for the semimonthly 

exchange cycle were used in this document because:

• A semimonthly exchange cycle was applied to missed external dose for this period (see 

Table C-1).

• The neutron-to-photon ratio values for semimonthly exchange were equal to or greater than 

the monthly exchange values.
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ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Table C-2. Lognormal neutron-to-photon ratio values, 1970 to 1976. 

Year

Neutron-to-photon ratio 

GM (semimonthly 

exchange) GSD

95th 

percentile

Neutron-to-photon 

ratio GM (monthly 

exchange) GSD

95th 

percentile

1970 1.61a 3.45 12.4 N/Ab N/A N/A

1971 1.61 3.45 12.4 N/A N/A N/A

1972 1.32 2.15 4.64 0.8 2.63 3.94

1973 1.32c 2.15 4.64 N/A N/A N/A

1974 0.68 3.01 4.16 N/A N/A N/A

1975 0.67 3.31 4.82 N/A N/A N/A

1976 0.95 3.59 7.81 0.78 4.29 8.55

a. Data for 1970 were not available. This value is the greater of the ratios for 1969 and 1971. The high neutron-to-gamma 

ratio in 1971 for Building 771 is reasonable from a process aspect, and extrapolating that ratio back to 1970 also is 

reasonable. In the aftermath of the 1969 plutonium fire in Buildings 776 and 777, the salvaged plutonium oxide had to 

be converted back to plutonium metal, a process done in Building 771. Building 771 had a huge backlog for relatively 

pure plutonium (little americium, low gamma) to be reprocessed, which was staged in the 776 to 771 tunnel and any 

other possible staging area in the vicinity. In addition, PuF4 (a high neutron source) seemed to be generated faster in 

Building 771 than it could be reduced to metal, which also caused a staging problem in or near the process areas in 

Building 771 until the backlog could be reduced.

b. N/A = not applicable.

c. Data for 1973 were not available. This value is the greater of the ratios for 1972 and 1974.

Table C-2 lists the neutron-to-photon ratios for this period. The individual ratios for available data 

were analyzed, and they fit lognormal distributions. 

As described below, an approach favorable to claimants was adopted in the development of coworker 

dose summaries; this approach was intended to account for any underestimation of doses to 

radiological workers at the RFP based on these considerations.

C.6 COWORKER ANNUAL DOSE SUMMARIES

Based on the described information and approaches, RFP coworker annual external dosimetry 

summaries were developed for use in the evaluation of external penetrating and nonpenetrating dose 

for certain workers who were potentially exposed to workplace radiation but for whom there is no or 

limited monitoring data from DOE. These summaries were developed using the following steps:

C.7 PLUTONIUM WORKERS

C.7.1 1952 to 1970

Step 1. As described above, for data from between 1952 and 1970, dose is based on a daily dose 

rate assigned continuously for every day of the year. 

Step 2. When the calculated dose rate was determined to be zero, missed dose was assigned for the 

entire year. The average exchange rate was determined based on recorded cycle exchange 

rates for the given building and year. This exchange rate was then used to estimate the 

expected number of badge exchanges for a year.

Step 3. The 50th- and 95th-percentile annual coworker gamma and neutron doses were derived from 

the doses calculated in Step 1 and 2.

Step 4. Table C-3 lists the results of the coworker analysis of the NDRP data. These percentile 

doses should be used for RFP plutonium workers with no or limited monitoring data through 
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ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

the use of the methodologies in Section 7.0 of ORAUT-OTIB-0020 (ORAUT 2011). In 

general, the 50th-percentile dose can be used as a best estimate of a worker’s dose when 

professional judgment indicates that the worker was probably exposed to intermittent low 

levels of external radiation. The 50th-percentile dose should generally not be used for 

workers who were routinely exposed. For routinely exposed workers (i.e., workers who were 

expected to have been monitored and routinely exposed), the 95th-percentile dose should be 

applied. However, other options are available through the guidance in ORAUT-OTIB-0020. 

For instance, for cases in which routine monitoring data exist and coworker dose is used to 

supplement missing quarters or years, the percentile dose should be the one that is 

consistent with the recorded doses unless there is reason to believe that the worker’s job or 

location differs significantly from that held during the year in which the dose was recorded. 

For workers who are unlikely to have been exposed, external onsite ambient dose should be 

used rather than coworker doses. 

Step 5. For instances where there was no building-specific data, the annual dose was based on the 

higher of the two adjacent years. 

Step 6. In instances where nonpenetrating dose is needed for dose estimation, Table C-5 values 

should be used. 

Step 7. Table C-6 lists the penetrating dose values (as described in the steps above) that have been 

adjusted using the guidance in Section 8.0 of ORAUT-OTIB-0052, Parameters to Consider 

When Processing Claims for Construction Trade Workers (ORAUT 2014). This guidance is 

applicable for construction trade workers who meet the criteria in Section 3.0 of that 

document. Because ORAUT-OTIB-0052 does not provide an adjustment factor for 

nonpenetrating dose, the guidance in Step 6 should be used to derive the nonpenetrating 

dose component for construction trade workers.

C.7.2 1971 to 2005

Step 1. As described above, for data from 1971 to 2005, the reported penetrating dose, which 

represents annual summary data, was modified for each worker to account for partial years of 

employment. This adjustment was made by analyzing the dosimetry wear dates in the HIS20 

database. For example, if the average employment period for all RFP employees in the 

NIOSH-Division of Compensation Analysis and Support Claims Tracking System (NOCTS) 

was 11 months in a particular calendar year, the reported annual doses were multiplied by 

12/11 (1.09). This permits the dose reconstructor to assign an appropriate prorated dose to 

account for partial years of employment or potential exposure.

Step 2. One-half of the maximum potential annual missed doses listed in Table C-1 were added to 

the reported annual doses from Step 1 (with the exception of reported positive doses, in 

which case the maximum missed dose was reduced by the dose that corresponded to one 

badge exchange, because it is not possible that all individual badge results were zero if a 

positive annual dose was reported).

Step 3. The 50th- and 95th-percentile annual coworker gamma doses were derived from the doses 

from Step 2 by ranking the data into cumulative probability curves and extracting the 50thand 95th-percentile doses for each year.
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ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Step 4. Table C-4 lists the results of the coworker analysis. These percentile doses should be used 

for RFP workers with no or limited monitoring data through the use of the methodologies in 

Section 7.0 of ORAUT-OTIB-0020 (ORAUT 2011). In general, the 50th-percentile dose can 

be used as a best estimate of a worker’s dose when professional judgment indicates that the 

worker was probably exposed to intermittent low levels of external radiation. The 50thpercentile dose should generally not be used for workers who were routinely exposed. For 

routinely exposed workers (i.e., workers who were expected to have been monitored and 

routinely exposed), the 95th-percentile dose should be applied. However, other options are 

available through the guidance in ORAUT-OTIB-0020. For instance, for cases in which 

routine monitoring data exist and coworker dose is used to supplement missing quarters or 

years, the percentile dose should be the one that is consistent with the recorded doses unless 

there is reason to believe that the worker’s job or location differs significantly from that held 

during the year in which the dose was recorded. For workers who are unlikely to have been 

exposed, external onsite ambient dose should be used rather than coworker doses.

Step 5. Neutron and gamma dose component values are listed in Table C-4. Neutron and gamma 

components of the penetrating dose for 1970 to 1976 were separated using the neutron-tophoton ratio values in Section 6.2 of this document. Neutron and gamma components of the 

penetrating dose for 1977 to 2005 were separated using neutron-to-photon ratio values.

Step 6. In instances where nonpenetrating dose is needed for dose estimation, Table C-5 values 

should be used. 

Step 7. Table C-7 lists penetrating dose values (as described in the steps above) that have been 

adjusted using the guidance in Section 8.0 of ORAUT-OTIB-0052 (ORAUT 2014). This 

guidance is applicable for construction trade workers who meet the criteria in Section 3.0 of 

that document. Because ORAUT-OTIB-0052 does not provide an adjustment factor for 

nonpenetrating dose, the guidance in Step 6 should be used to derive the nonpenetrating 

dose component for construction trade workers.

C.8 URANIUM WORKERS

Step 1. As described above, for data from 1952 to 2005, the reported penetrating dose, which 

represented annual summary data, was modified for each worker to account for partial years 

of employment. This adjustment was made by analyzing the dosimetry wear dates in the 

HIS20 database. For example, if the average employment period for all RFP employees in 

NOCTS was 11 months in a particular calendar year, the reported annual doses were 

multiplied by 12/11 (1.09). This permits the dose reconstructor to assign an appropriate 

prorated dose to account for partial years of employment or potential exposure.

Step 2. One-half of the maximum potential annual missed doses listed in Table C-1 was added to the 

reported annual doses from Step 1 (with the exception of reported positive doses, in which 

case the maximum missed dose was reduced by the dose that corresponded to one badge 

exchange because it is not possible that all individual badge results were zero if a positive 

annual dose was reported).

Step 3. The 50th- and 95th-percentile annual coworker gamma doses were derived from the doses 

from Step 2 by ranking the data into cumulative probability curves and extracting the 50thand 95th-percentile doses for each year.
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ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Step 4. Table C-5 lists the results of the coworker analysis. These percentile doses should be used 

for RFP workers with no or limited monitoring data through the use of the methodologies in 

Section 7.0 of ORAUT-OTIB-0020 (ORAUT 2011). In general, the 50th-percentile dose can 

be used as a best estimate of a worker’s dose when professional judgment indicates that the 

worker was probably exposed to intermittent low levels of external radiation. The 

50th-percentile dose should generally not be used for workers who were routinely exposed. 

For routinely exposed workers (i.e., workers who were expected to have been monitored and 

routinely exposed), the 95th-percentile dose should be applied. However, other options are 

available through the guidance in ORAUT-OTIB-0020. For instance, for cases in which 

routine monitoring data exist and coworker dose is used to supplement missing quarters or 

years, the percentile dose should be the one that is consistent with the recorded doses unless 

there is reason to believe that the worker’s job or location differs significantly from that held 

during the years dose was recorded. For workers who are unlikely to have been exposed, 

external onsite ambient dose should be used rather than coworker doses. Also, for the 1969 

to 1970 time period, the data from the HIS-20 database were analyzed without consideration 

of the zero readings due to issues regarding the accurate recording of “zero” dose readings 

during that period. 

Step 5. In instances where nonpenetrating dose is needed for dose estimation, Table C-5 values 

should be used. 

Step 6. Table C-8 lists penetrating dose values (as described in the steps above) that have been 

adjusted using the guidance in Section 8.0 of ORAUT-OTIB-0052 (ORAUT 2014). This 

guidance is applicable for construction trade workers who meet the criteria in Section 3.0 of 

that document. Because ORAUT-OTIB-0052 does not provide an adjustment factor for 

nonpenetrating dose, the guidance in Step 5 should be used to derive the nonpenetrating 

dose component for construction trade workers.
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ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

C.9 COWORKER DOSE TABLES 

Table C-3. Annual external coworker doses for plutonium workers, 1952 to 1970 (rem/yr). 

Building 71

Year

Photon 

50th percentile

Photon 

95th percentile

Neutron 

50th percentile

Neutron 

95th percentile

1952 Not applicable Not applicable Not applicable Not applicable

1953 0.216 1.752 Not applicable Not applicable

1954 0.520a 3.340 Not applicable Not applicable

1955 0.469 6.518 Not applicable Not applicable

1956 0.456 4.890 Not applicable Not applicable

1957 0.562 5.371 Not applicable Not applicable

1958 0.556 6.309 Not applicable Not applicable

1959 1.356 6.779 Not applicable Not applicable

1960 0.574 3.546 Not applicable Not applicable

1961 1.330 4.849 Not applicable Not applicable

1962 1.069 4.105 Not applicable Not applicable

1963 0.574 3.337 Not applicable Not applicable

1964 0.261 2.320 Not applicable Not applicable

1965 0.523 4.000 Not applicable Not applicable

1966 1.192 6.037 Not applicable Not applicable

1967 0.907 5.614 0.808 3.941

1968 0.678 3.468 0.991 6.935

1969 0.590 4.278 1.590 4.824

1970 0.120a 4.278c 1.675 5.110

Building 76

Year

Photon 

50th percentile

Photon 

95th percentile

Neutron 

50th percentile

Neutron 

95th percentile

1952 Not applicable Not applicable Not applicable Not applicable

1953 Not applicable Not applicable Not applicable Not applicable

1954 Not applicable Not applicable Not applicable Not applicable

1955 Not applicable Not applicable Not applicable Not applicable

1956 Not applicable Not applicable Not applicable Not applicable

1957 0.520a 2.699 Not applicable Not applicable

1958 0.730 6.361 Not applicable Not applicable

1959 1.616 6.153 Not applicable Not applicable

1960 1.981 5.840 Not applicable Not applicable

1961 2.451 5.319 Not applicable Not applicable

1962 1.721 4.484 Not applicable Not applicable

1963 1.616 4.563 Not applicable Not applicable

1964 1.741 5.079 Not applicable Not applicable

1965 0.913 5.342 Not applicable Not applicable

1966 1.898 7.324 Not applicable Not applicable

1967 0.706 6.486 0.971 2.774

1968 0.283 2.759 0.954 3.441

1969 0.065 1.570 1.217 2.292

1970 0.120a 1.102 1.703 2.738
Document No. ORAUT-TKBS-0011-6 Revision No. 03 Effective Date: 01/14/2019 Page 102 of 109

ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Building 77

Year

Photon 

50th percentile

Photon 

95th percentile

Neutron 

50th percentile

Neutron 

95th percentile

1952 Not applicable Not applicable Not applicable Not applicable

1953 Not applicable Not applicable Not applicable Not applicable

1954 Not applicable Not applicable Not applicable Not applicable

1955 Not applicable Not applicable Not applicable Not applicable

1956 Not applicable Not applicable Not applicable Not applicable

1957 0.480a 0.480a Not applicable Not applicable

1958 0.052 2.918 Not applicable Not applicable

1959 0.261 2.868 Not applicable Not applicable

1960 0.130 3.598 Not applicable Not applicable

1961 1.121 5.214 Not applicable Not applicable

1962 0.479 2.764 Not applicable Not applicable

1963 0.815 4.343 Not applicable Not applicable

1964 0.965 4.958 Not applicable Not applicable

1965 0.754 3.682 Not applicable Not applicable

1966 1.278 5.519 Not applicable Not applicable

1967 1.454 5.299 1.048 1.966

1968 0.586 2.772 0.969 1.811

1969 0.088 1.991 0.717 1.325

1970 0.120a 2.392 2.060 2.873

Building 91

Year

Photon 

50th percentile

Photon 

95th percentile

Neutron 

50th percentile

Neutron 

95th percentile

1952 0.782 1.646 Not applicable Not applicable

1953 0.332 2.257 Not applicable Not applicable

1954 0.520a 0.782 Not applicable Not applicable

1955 0.520a 0.913 Not applicable Not applicable

1956 0.480a 0.521 Not applicable Not applicable

1957 0.520a 0.521 Not applicable Not applicable

1958 0.480a 0.521 Not applicable Not applicable

1959 0.240a 0.521 Not applicable Not applicable

1960 0.240a 0.261 Not applicable Not applicable

1961 0.240a 0.521 Not applicable Not applicable

1962 0.240a 0.519 Not applicable Not applicable

1963 0.240a 0.587 Not applicable Not applicable

1964 0.240a 0.417 Not applicable Not applicable

1965 0.240a 0.706 Not applicable Not applicable

1966 0.118 1.662 Not applicable Not applicable

1967 0.240a 1.276 1.189b 1.669b

1968 0.240a 0.665 0.669 1.073

1969 0.120a 0.483 1.108 2.237

1970 0.209 1.322 1.108b 2.237b
Document No. ORAUT-TKBS-0011-6 Revision No. 03 Effective Date: 01/14/2019 Page 103 of 109

ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

All Buildings

Year

Photon 

50th percentile

Photon 

95th percentile

Neutron 

50th percentile

Neutron 

95th percentile

1952 0.782 1.646 Not applicable Not applicable

1953 0.261 1.947 Not applicable Not applicable

1954 0.520a 1.825 Not applicable Not applicable

1955 0.240a 4.596 Not applicable Not applicable

1956 0.130 3.859 Not applicable Not applicable

1957 0.480a 3.702 Not applicable Not applicable

1958 0.140 4.355 Not applicable Not applicable

1959 0.782 5.788 Not applicable Not applicable

1960 0.433 4.745 Not applicable Not applicable

1961 1.051 5.058 Not applicable Not applicable

1962 0.730 3.964 Not applicable Not applicable

1963 0.548 3.807 Not applicable Not applicable

1964 0.339 3.285 Not applicable Not applicable

1965 0.393 3.852 Not applicable Not applicable

1966 0.900 5.958 Not applicable Not applicable

1967 0.482 5.137 0.873 3.134

1968 0.281 2.869 0.965 5.044

1969 0.135 2.997 1.348 4.356

1970 0.120a 1.200 1.314 4.248

a. Dose rate was determined to be zero; missed dose was applied based on the number of exchange 

cycles observed for that building.

b. Dose rate was based on the favorable to claimant dose rate associated with higher of adjacent year.

Table C-4. Annual external coworker doses for plutonium workers, 1971 to 2005 (rem). 

Year

Photon

95th percentile

Photon

50th percentile

Neutron

95th percentile

Neutron

50th percentile

Number of 

monitored workers

1971 0.743 0.147 1.196 0.237 3,398

1972 0.799 0.163 1.054 0.215 3,282

1973 0.731 0.164 0.966 0.216 3,020

1974 1.119 0.296 0.761 0.201 2,687

1975 0.796 0.240 0.533 0.161 2,489

1976 0.424 0.127 0.402 0.121 2,424

1977 0.415 0.086 0.174 0.036 3,740

1978 0.492 0.085 0.207 0.035 4,176

1979 0.638 0.086 0.268 0.036 3,893

1980 0.523 0.085 0.220 0.035 3,752

1981 0.600 0.077 0.252 0.033 4,060

1982 0.697 0.085 0.293 0.036 4,851

1983 0.854 0.094 0.359 0.040 5,360

1984 0.848 0.100 0.356 0.042 5,673

1985 0.917 0.100 0.385 0.042 6,140

1986 0.961 0.129 0.404 0.054 4,942

1987 1.063 0.182 0.447 0.077 2,583

1988 0.870 0.140 0.366 0.059 2,778

1989 0.388 0.081 0.163 0.034 5,296

1990 0.203 0.091 0.085 0.038 3,369

1991 0.237 0.108 0.100 0.045 5,641

1992 0.188 0.105 0.079 0.044 5,831
Document No. ORAUT-TKBS-0011-6 Revision No. 03 Effective Date: 01/14/2019 Page 104 of 109

ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Year

Photon

95th percentile

Photon

50th percentile

Neutron

95th percentile

Neutron

50th percentile

Number of 

monitored workers

1993 0.146 0.056 0.061 0.024 5,313

1994 0.126 0.059 0.053 0.025 4,839

1995 0.141 0.056 0.059 0.024 4,130

1996 0.193 0.074 0.081 0.031 3,454

1997 0.248 0.060 0.104 0.025 3,718

1998 0.193 0.051 0.081 0.022 3,470

1999 0.136 0.050 0.057 0.021 3,655

2000 0.115 0.042 0.048 0.018 3,576

2001 0.112 0.042 0.047 0.018 3,443

2002 0.110 0.042 0.046 0.018 3,502

2003 0.089 0.042 0.038 0.018 3,373

2004 0.068 0.042 0.029 0.018 2,758

2005 0.123 0.050 0.052 0.021 955

Table C-5. Annual external coworker doses for uranium workers, 1952 to 2005 (rem). 

Year

Penetrating, 

95th percentile

Penetrating, 

50th percentile

Nonpenetrating, 

95th percentile

Nonpenetrating, 

50th percentile

Number of 

monitored workers

1952 5.018 2.505 5.133 2.620 42

1953 4.190 0.751 4.553 0.892 319

1954 3.233 1.095 3.600 1.361 353

1955 4.411 1.165 5.266 1.431 529

1956 4.461 1.135 5.617 1.415 781

1957 5.136 1.177 6.004 1.454 918

1958 6.015 1.253 7.553 1.584 1,062

1959 7.186 1.581 8.002 1.908 1,063

1960 7.121 1.293 7.728 1.645 1,284

1961 7.850 1.527 8.201 1.923 1,638

1962 6.523 1.542 7.062 1.828 2,003

1963 5.955 0.940 6.232 1.104 2,176

1964 4.875 0.648 5.012 0.799 2,834

1965 3.533 0.511 3.663 0.598 2,826

1966 4.767 0.592 4.976 0.679 2,888

1967 4.379 0.627 4.735 0.761 2,902

1968 3.276 0.578 3.591 0.714 3,101

1969 3.588 0.680 3.788 0.824 3,471

1970 2.894 0.531 3.067 0.651 3,308

1971 1.938 0.384 2.096 0.500 3,398

1972 1.853 0.377 1.995 0.494 3,282

1973 1.697 0.380 1.848 0.497 3,020

1974 1.881 0.497 2.047 0.612 2,687

1975 1.329 0.401 1.525 0.516 2,489

1976 0.826 0.248 1.030 0.364 2,424

1977 0.589 0.122 0.699 0.177 3,740

1978 0.698 0.120 0.830 0.180 4,176

1979 0.906 0.122 1.074 0.180 3,893

1980 0.743 0.120 0.889 0.168 3,752

1981 0.853 0.110 1.008 0.180 4,060

1982 0.990 0.121 1.174 0.166 4,851

1983 1.212 0.134 1.440 0.193 5,360

1984 1.204 0.141 1.551 0.200 5,673
Document No. ORAUT-TKBS-0011-6 Revision No. 03 Effective Date: 01/14/2019 Page 105 of 109

ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Year

Penetrating, 

95th percentile

Penetrating, 

50th percentile

Nonpenetrating, 

95th percentile

Nonpenetrating, 

50th percentile

Number of 

monitored workers

1985 1.302 0.142 1.584 0.203 6,140

1986 1.365 0.183 1.860 0.301 4,942

1987 1.510 0.259 2.265 0.457 2,583

1988 1.236 0.199 1.614 0.373 2,778

1989 0.550 0.115 0.733 0.180 5,296

1990 0.288 0.130 0.435 0.201 3,369

1991 0.337 0.154 0.471 0.228 5,641

1992 0.267 0.150 0.421 0.235 5,831

1993 0.207 0.080 0.296 0.150 5,313

1994 0.179 0.084 0.242 0.142 4,839

1995 0.200 0.080 0.275 0.137 4,130

1996 0.274 0.105 0.357 0.168 3,454

1997 0.352 0.085 0.420 0.145 3,718

1998 0.275 0.073 0.337 0.130 3,470

1999 0.192 0.072 0.256 0.127 3,655

2000 0.164 0.060 0.224 0.120 3,576

2001 0.160 0.060 0.223 0.120 3,443

2002 0.157 0.060 0.215 0.120 3,502

2003 0.127 0.060 0.194 0.120 3,373

2004 0.097 0.060 0.164 0.120 2,758

2005 0.175 0.071 0.234 0.129 955

Table C-6. Annual external coworker doses for plutonium workers, 1952 to 1970, modified in 

accordance with ORAUT-OTIB-0052 (rem/yr) (ORAUT 2014).

Building 71

Year

Photon 

50th percentile

Photon 

95th percentile

Neutron 

50th percentile

Neutron 

95th percentile

1952 Not applicable Not applicable Not applicable Not applicable

1953 0.303 2.453 Not applicable Not applicable

1954 0.728a 4.676 Not applicable Not applicable

1955 0.657 9.125 Not applicable Not applicable

1956 0.639 6.846 Not applicable Not applicable

1957 0.786 7.519 Not applicable Not applicable

1958 0.779 8.833 Not applicable Not applicable

1959 1.898 9.490 Not applicable Not applicable

1960 0.803 4.964 Not applicable Not applicable

1961 1.862 6.789 Not applicable Not applicable

1962 1.497 5.747 Not applicable Not applicable

1963 0.803 4.672 Not applicable Not applicable

1964 0.365 3.249 Not applicable Not applicable

1965 0.732 5.600 Not applicable Not applicable

1966 1.669 8.451 Not applicable Not applicable

1967 1.269 7.860 1.132 5.518

1968 0.949 4.855 1.387 9.709

1969 0.825 5.989 2.227 6.754

1970 0.168a 5.989b 2.345 7.154
Document No. ORAUT-TKBS-0011-6 Revision No. 03 Effective Date: 01/14/2019 Page 106 of 109

ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Building 76

Year

Photon 

50th-percentile

Photon 

95th-percentile

Neutron 

50th-percentile

Neutron 

95th-percentile

1952 Not applicable Not applicable Not applicable Not applicable

1953 Not applicable Not applicable Not applicable Not applicable

1954 Not applicable Not applicable Not applicable Not applicable

1955 Not applicable Not applicable Not applicable Not applicable

1956 Not applicable Not applicable Not applicable Not applicable

1957 0.728a 3.778 Not applicable Not applicable

1958 1.022 8.906 Not applicable Not applicable

1959 2.263 8.614 Not applicable Not applicable

1960 2.774 8.176 Not applicable Not applicable

1961 3.431 7.446 Not applicable Not applicable

1962 2.409 6.278 Not applicable Not applicable

1963 2.263 6.388 Not applicable Not applicable

1964 2.437 7.110 Not applicable Not applicable

1965 1.278 7.479 Not applicable Not applicable

1966 2.657 10.254 Not applicable Not applicable

1967 0.989 9.081 1.360 3.884

1968 0.396 3.863 1.335 4.818

1969 0.091 2.197 1.703 3.209

1970 0.168a 1.543 2.385 3.833

Building 77

Year

Photon 

50th-percentile

Photon 

95th-percentile

Neutron 

50th-percentile

Neutron 

95th-percentile

1952 Not applicable Not applicable Not applicable Not applicable

1953 Not applicable Not applicable Not applicable Not applicable

1954 Not applicable Not applicable Not applicable Not applicable

1955 Not applicable Not applicable Not applicable Not applicable

1956 Not applicable Not applicable Not applicable Not applicable

1957 0.672a 0.672a Not applicable Not applicable

1958 0.073 4.086 Not applicable Not applicable

1959 0.365 4.015 Not applicable Not applicable

1960 0.183 5.037 Not applicable Not applicable

1961 1.570 7.300 Not applicable Not applicable

1962 0.670 3.869 Not applicable Not applicable

1963 1.141 6.080 Not applicable Not applicable

1964 1.351 6.942 Not applicable Not applicable

1965 1.055 5.155 Not applicable Not applicable

1966 1.789 7.726 Not applicable Not applicable

1967 2.036 7.419 1.467 2.753

1968 0.820 3.881 1.357 2.535

1969 0.123 2.787 1.004 1.855

1970 0.168a 3.349 2.885 4.022

Building 91

Year

Photon 

50th-percentile

Photon 

95th-percentile

Neutron 

50th-percentile

Neutron 

95th-percentile

1952 1.095 2.304 Not applicable Not applicable

1953 0.465 3.160 Not applicable Not applicable

1954 0.728a 1.095 Not applicable Not applicable

1955 0.728a 1.278 Not applicable Not applicable
Document No. ORAUT-TKBS-0011-6 Revision No. 03 Effective Date: 01/14/2019 Page 107 of 109

ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Year

Photon 

50th-percentile

Photon 

95th-percentile

Neutron 

50th-percentile

Neutron 

95th-percentile

1956 0.672a 0.730 Not applicable Not applicable

1957 0.728a 0.730 Not applicable Not applicable

1958 0.672a 0.730 Not applicable Not applicable

1959 0.336a 0.730 Not applicable Not applicable

1960 0.336a 0.365 Not applicable Not applicable

1961 0.336a 0.730 Not applicable Not applicable

1962 0.336a 0.727 Not applicable Not applicable

1963 0.336a 0.821 Not applicable Not applicable

1964 0.336a 0.584 Not applicable Not applicable

1965 0.336a 0.989 Not applicable Not applicable

1966 0.165 2.327 Not applicable Not applicable

1967 0.336a 1.786 1.665b 2.337b

1968 0.336a 0.931 0.936 1.503

1969 0.168a 0.676 1.551 3.131

1970 0.292 1.850 1.551b 3.132b

All buildings

Year

Photon 

50th-percentile

Photon 

95th-percentile

Neutron 

50th-percentile

Neutron 

95th-percentile

1952 1.095 2.304 Not applicable Not applicable

1953 0.365 2.725 Not applicable Not applicable

1954 0.728a 2.555 Not applicable Not applicable

1955 0.336a 6.435 Not applicable Not applicable

1956 0.183 5.402 Not applicable Not applicable

1957 0.672a 5.183 Not applicable Not applicable

1958 0.197 6.097 Not applicable Not applicable

1959 1.095 8.103 Not applicable Not applicable

1960 0.606 6.643 Not applicable Not applicable

1961 1.472 7.081 Not applicable Not applicable

1962 1.022 5.549 Not applicable Not applicable

1963 0.767 5.330 Not applicable Not applicable

1964 0.475 4.599 Not applicable Not applicable

1965 0.551 5.393 Not applicable Not applicable

1966 1.260 8.341 Not applicable Not applicable

1967 0.675 7.192 1.223 4.388

1968 0.393 4.016 1.352 7.061

1969 0.189 4.195 1.887 6.098

1970 0.168a 1.680 1.840 5.947

a. Dose rate was determined to be zero; missed dose was applied based on the number of exchange 

cycles observed for that building.

b. Dose rate was based on the favorable to claimant dose rate associated with higher of adjacent year.

Table C-7. Annual external coworker doses for plutonium workers, 1971 to 2005, modified in 

accordance with ORAUT-OTIB-0052 (rem) (ORAUT 2014).

Year

Photon

95th-percentile

Photon

50th-percentile

Neutron

95th-percentile

Neutron

50th-percentile

Number of 

monitored workers

1971 1.004 0.171 1.617 0.275 3,398

1972 1.079 0.188 1.424 0.248 3,282

1973 0.984 0.190 1.299 0.251 3,020

1974 1.512 0.359 1.028 0.244 2,687

1975 1.059 0.281 0.710 0.189 2,489
Document No. ORAUT-TKBS-0011-6 Revision No. 03 Effective Date: 01/14/2019 Page 108 of 109

ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Year

Photon

95th-percentile

Photon

50th-percentile

Neutron

95th-percentile

Neutron

50th-percentile

Number of 

monitored workers

1976 0.546 0.131 0.519 0.124 2,424

1977 0.550 0.089 0.231 0.037 3,740

1978 0.658 0.085 0.276 0.035 4,176

1979 0.862 0.089 0.362 0.037 3,893

1980 0.702 0.085 0.295 0.035 3,752

1981 0.810 0.077 0.340 0.033 4,060

1982 0.945 0.086 0.397 0.036 4,851

1983 1.164 0.101 0.489 0.042 5,360

1984 1.156 0.108 0.486 0.046 5,673

1985 1.253 0.109 0.526 0.046 6,140

1986 1.315 0.149 0.552 0.063 4,942

1987 1.458 0.225 0.612 0.094 2,583

1988 1.187 0.165 0.499 0.069 2,778

1989 0.512 0.082 0.215 0.035 5,296

1990 0.253 0.097 0.106 0.041 3,369

1991 0.301 0.121 0.126 0.051 5,641

1992 0.232 0.116 0.098 0.049 5,831

1993 0.188 0.063 0.079 0.027 5,313

1994 0.161 0.067 0.068 0.028 4,839

1995 0.182 0.063 0.076 0.027 4,130

1996 0.255 0.088 0.107 0.037 3,454

1997 0.332 0.069 0.139 0.029 3,718

1998 0.255 0.057 0.107 0.024 3,470

1999 0.174 0.055 0.073 0.023 3,655

2000 0.146 0.042 0.061 0.018 3,576

2001 0.142 0.042 0.060 0.018 3,443

2002 0.139 0.042 0.058 0.018 3,502

2003 0.110 0.042 0.046 0.018 3,373

2004 0.080 0.042 0.034 0.018 2,758

2005 0.157 0.054 0.066 0.023 955

Table C-8. Annual external coworker doses for uranium workers, 1952 to 2005, modified in 

accordance with ORAUT-OTIB-0052 (rem) (ORAUT 2014).

Year

Penetrating

95th percentile

Penetrating

50th percentile

Nonpenetrating

95th percentile

Nonpenetrating

50th percentile

Number of 

monitored workers

1952 6.842 3.323 5.133 2.620 42

1953 5.681 0.867 4.553 0.892 319

1954 4.118 1.125 3.600 1.361 353

1955 5.768 1.224 5.266 1.431 529

1956 5.838 1.181 5.617 1.415 781

1957 6.782 1.240 6.004 1.454 918

1958 8.012 1.346 7.553 1.584 1,062

1959 9.653 1.805 8.002 1.908 1,063

1960 9.561 1.402 7.728 1.645 1,284

1961 10.583 1.730 8.201 1.923 1,638

1962 8.724 1.751 7.062 1.828 2,003

1963 8.153 1.132 6.232 1.104 2,176

1964 6.641 0.723 5.012 0.799 2,834

1965 4.858 0.627 3.663 0.598 2,826

1966 6.585 0.740 4.976 0.679 2,888
Document No. ORAUT-TKBS-0011-6 Revision No. 03 Effective Date: 01/14/2019 Page 109 of 109

ATTACHMENT C 

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT (continued)

Year

Penetrating

95th percentile

Penetrating

50th percentile

Nonpenetrating

95th percentile

Nonpenetrating

50th percentile

Number of 

monitored workers

1967 5.947 0.694 4.735 0.761 2,902

1968 4.402 0.625 3.591 0.714 3,101

1969 4.931 0.860 3.788 0.824 3,471

1970 3.959 0.652 3.067 0.651 3,308

1971 2.622 0.446 2.096 0.500 3,398

1972 2.503 0.436 1.995 0.494 3,282

1973 2.284 0.441 1.848 0.497 3,020

1974 2.541 0.603 2.047 0.612 2,687

1975 1.769 0.470 1.525 0.516 2,489

1976 1.064 0.255 1.030 0.364 2,424

1977 0.781 0.127 0.699 0.177 3,740

1978 0.934 0.120 0.830 0.180 4,176

1979 1.224 0.127 1.074 0.180 3,893

1980 0.996 0.120 0.889 0.168 3,752

1981 1.150 0.110 1.008 0.180 4,060

1982 1.343 0.122 1.174 0.166 4,851

1983 1.653 0.143 1.440 0.193 5,360

1984 1.642 0.154 1.551 0.200 5,673

1985 1.779 0.154 1.584 0.203 6,140

1986 1.867 0.212 1.860 0.301 4,942

1987 2.070 0.319 2.265 0.457 2,583

1988 1.686 0.234 1.614 0.373 2,778

1989 0.727 0.117 0.733 0.180 5,296

1990 0.360 0.137 0.435 0.201 3,369

1991 0.427 0.171 0.471 0.228 5,641

1992 0.330 0.165 0.421 0.235 5,831

1993 0.267 0.090 0.296 0.150 5,313

1994 0.229 0.095 0.242 0.142 4,839

1995 0.258 0.090 0.275 0.137 4,130

1996 0.362 0.124 0.357 0.168 3,454

1997 0.471 0.098 0.420 0.145 3,718

1998 0.363 0.080 0.337 0.130 3,470

1999 0.247 0.078 0.256 0.127 3,655

2000 0.207 0.060 0.224 0.120 3,576

2001 0.202 0.060 0.223 0.120 3,443

2002 0.197 0.060 0.215 0.120 3,502

2003 0.156 0.060 0.194 0.120 3,373

2004 0.114 0.060 0.164 0.120 2,758

2005 0.223 0.077 0.234 0.129 955