ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 105

DOE Review Release 11/03/2010

Document Title:

Rocky Flats Plant – Occupational External Dose

Document Number: ORAUT-TKBS-0011-6 

Revision: 02 PC-1 

Effective Date: 10/20/2010

Type of Document: TBD

Supersedes: Revision 02 

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

Site Expert(s): James M. Langsted

Approval: Signature on File Approval Date: 10/07/2010

Mutty M. Sharfi, Document Owner

Concurrence: Signature on File Concurrence Date: 10/07/2010

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 10/11/2010

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 10/11/2010

James P. Griffin, Deputy Project Director

Approval: Signature on File Approval Date: 10/20/2010

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 2 of 105

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.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 3 of 105

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

8

6.1 Introduction ............................................................................................................................10

6.1.1 Purpose......................................................................................................................11

6.1.2 Scope.........................................................................................................................11

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

6.3 Interpreting the External Dosimetry Record............................................................................ 16

6.3.1 Dosimetry Records Systems.......................................................................................16

6.3.2 Observed Data Discrepancies ....................................................................................16

6.3.2.1 Rounding ..................................................................................................... 17

6.3.2.2 Deep Dose Not Equal to Gamma Plus Neutron Doses................................. 17

6.3.2.2.1 Possible Algorithm Issue ............................................................ 17

6.3.2.2.2 Possible Manual Correction........................................................ 17

6.3.3 Database Table-Specific Issues .................................................................................17

6.3.3.1 RHRST_ED_TLD_HISTORY ....................................................................... 18

6.3.3.1.1 1976 Records (Individual Employed After 1976)......................... 18

6.3.3.1.2 1976 Records (Individual Employed Before 1976)...................... 18

6.3.3.1.3 Post-1976 Records..................................................................... 18

6.3.3.2 RHRST_ED_TLD_DOS ............................................................................... 18

6.3.3.2.1 Dose History Hard-Copy File Contents....................................... 19

6.3.4 Interpretation of Dosimetry Data.................................................................................21

6.3.5 Additional Data Available............................................................................................21

6.3.5.1 Rocky Flats Work History File ...................................................................... 21

6.3.5.2 Neutron Dose Reconstruction Project File.................................................... 21

6.3.5.3 Job Exposure Matrix .................................................................................... 21

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 ...........................................................................................29

6.5 Common Issues .....................................................................................................................30

6.5.1 Number of Zero Readings ..........................................................................................30

6.5.2 Discrepancies.............................................................................................................30

6.5.3 Missing Entry..............................................................................................................32

6.5.4 Exposure Geometry....................................................................................................32

6.5.5 Lead Aprons...............................................................................................................32

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

6.5.7 Potential Elevated Background Subtraction................................................................34

6.5.8 Badge Reading Policy in 1969-1970...........................................................................35

6.6 Photon Dose ..........................................................................................................................35

6.6.1 Energy Groups ...........................................................................................................35

6.6.1.1 Default Exposure Spectra ............................................................................ 36

6.6.1.2 Dosimeter-Indicated Photon Energy............................................................. 37
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6.6.1.2.1 Pre-1960 .................................................................................... 37

6.6.1.2.2 1960–1970 ................................................................................. 37

6.6.1.2.3 1970–Present............................................................................. 38

6.6.2 Calibration Factor .......................................................................................................38

6.6.2.1 Reported-Dose-to-Organ-Dose Conversion Factor Units ............................. 38

6.6.3 Missed Dose...............................................................................................................39

6.6.3.1 Limit of Detection ......................................................................................... 39

6.6.3.2 Number of Zero Readings ............................................................................ 40

6.6.3.3 Determination of Missed Dose ..................................................................... 40

6.6.3.4 Unmonitored Energy Range......................................................................... 40

6.6.4 Geometry....................................................................................................................41

6.6.4.1 Angular Dependence ................................................................................... 41

6.6.4.2 Exposure Geometry ..................................................................................... 41

6.6.5 Uncertainty .................................................................................................................41

6.6.5.1 Film..............................................................................................................41

6.6.5.2 Thermoluminescent Dosimeter..................................................................... 42

6.6.5.2.1 Loose-Chip Thermoluminescent Dosimeters .............................. 42

6.6.5.2.2 Panasonic Thermoluminescent Dosimeter ................................. 43

6.7 Neutron Dose.........................................................................................................................44

6.7.1 Energy Groups ...........................................................................................................44

6.7.1.1 Exposure Spectra......................................................................................... 44

6.7.1.2 Reported Dose to Energy Groups ................................................................ 45

6.7.2 Calibration Factor .......................................................................................................45

6.7.2.1 Dosimeter-Specific Quality Factor Conversion ............................................. 45

6.7.2.2 Reported-Dose-to-Organ-Dose Conversion Factor Units ............................. 45

6.7.3 Missed Dose...............................................................................................................46

6.7.3.1 Limit of Detection ......................................................................................... 46

6.7.3.2 Number of Zero Readings ............................................................................ 47

6.7.3.3 Neutron Dose Reconstruction Project .......................................................... 47

6.7.3.4 Default Neutron-to-Gamma Ratio................................................................. 49

6.7.4 Geometry....................................................................................................................50

6.7.4.1 Angular Dependence ................................................................................... 50

6.7.4.2 Exposure Geometry ..................................................................................... 50

6.7.5 Uncertainty .................................................................................................................50

6.7.5.1 Film..............................................................................................................50

6.7.5.2 Thermoluminescent Dosimeter..................................................................... 51

6.8 Electron Dose ........................................................................................................................51

6.8.1 Energy Groups ...........................................................................................................52

6.8.1.1 Exposure Spectra......................................................................................... 52

6.8.1.2 Reported Dose to Energy Groups ................................................................ 53

6.8.2 Calibration Factor .......................................................................................................53

6.8.2.1 Reported-Dose-to-Organ-Dose Conversion Factor Units ............................. 53

6.8.3 Missed Dose...............................................................................................................53

6.8.3.1 Limit of Detection ......................................................................................... 53

6.8.3.2 Number of Zero Readings ............................................................................ 54

6.8.3.3 Unmonitored Energy Range......................................................................... 54

6.8.4 Geometry....................................................................................................................55

6.8.4.1 Angular Dependence ................................................................................... 55

6.8.4.2 Exposure Geometry ..................................................................................... 55

6.8.5 Uncertainty .................................................................................................................55

6.8.5.1 Film..............................................................................................................55
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6.8.5.2 Thermoluminescent Dosimeter..................................................................... 56

6.8.5.2.1 Loose-Chip Thermoluminescent Dosimeters .............................. 56

6.8.5.2.2 Panasonic TLD Dosimeter.......................................................... 56

6.8.6 Skin Contamination ....................................................................................................57

6.9 Unmonitored Individuals......................................................................................................... 58

6.9.1 In Production Areas ....................................................................................................58

6.9.2 Outside Production Areas...........................................................................................58

6.10 Extremity Dosimetry...............................................................................................................58

6.11 Attributions and Annotations .................................................................................................. 59

References ........................................................................................................................................63

Glossary ............................................................................................................................................67

ATTACHMENT A, EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS........................... 71

ATTACHMENT B, MAJOR JOB CATEGORIES ................................................................................ 87

ATTACHMENT C, EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT ..... 89

C.1 General Approach..................................................................................................................90

C.2 Applications and Limitations................................................................................................... 90

C.3 Coworker Data Development ................................................................................................. 90

C.4 Adjustment for Missed Dose .................................................................................................. 91

C.4.1 Plutonium Workers (Before 1971)...............................................................................91

C.4.2 Plutonium (After 1970) and Uranium (all years) Workers............................................ 91

C.5 Special Considerations .......................................................................................................... 92

C.6 Coworker Annual Dose Summaries ....................................................................................... 93

C.7 Plutonium Workers................................................................................................................. 93

C.7.1 1952 to 1970 ..............................................................................................................93

C.7.2 1971 to 2005 ..............................................................................................................94

C.8 Uranium Workers ...................................................................................................................95

C.9 Coworker Dose Tables........................................................................................................... 97
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 6 of 105

LIST OF TABLES

TABLE TITLE

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

PAGE

14

6-2 Interpretation of reported data................................................................................................ 22

6-3 Conservatively determined default dosimeter exchange frequencies ..................................... 28

6-4 Summary of historical recorded dose practices ...................................................................... 29

6-5 Dose limits based on exchange frequency ............................................................................. 31

6-6 Exposure geometry calculation .............................................................................................. 32

6-7 Exposure geometry defaults for major job categories............................................................. 33

6-8 Bias correction factors for application to dose received while wearing a lead apron............... 33

6-9 Photon energy distributions.................................................................................................... 36

6-10 Default photon energy distributions ........................................................................................ 37

6-11 Photon dose units for use with organ dose conversion factors ............................................... 39

6-12 Photon LODs .........................................................................................................................40

6-13 Uncertainty for photon film dose............................................................................................. 42

6-14 Uncertainty for loose-chip TLD photon dose........................................................................... 43

6-15 Uncertainty for DOELAP-accredited TLD photon dose........................................................... 43

6-16 Neutron dose measurements divided into energy groups....................................................... 44

6-17 Default neutron energy distribution......................................................................................... 45

6-18 Neutron dose units for use with organ dose conversion factors.............................................. 46

6-19 Neutron film track counting detail ........................................................................................... 46

6-20 Neutron LODs........................................................................................................................47

6-21 RFP lognormal neutron-to-photon ratio values, 1970 to 1976 ................................................ 50

6-22 ORAU Team-developed neutron-to-gamma ratios ................................................................. 50

6-23 Uncertainty for TLD neutron dose measurements .................................................................. 51

6-24 Beta LODs .............................................................................................................................54

6-25 Uncertainty for beta film readings........................................................................................... 55

6-26 Uncertainty for loose-chip TLD beta dose .............................................................................. 56

6-27 Uncertainty for DOELAP-accredited TLD beta dose............................................................... 56

6-28 DU mixtures ...........................................................................................................................57

6-29 One-year-old DU....................................................................................................................57

C-1 Missed external doses ........................................................................................................... 92

C-2 Lognormal neutron-to-photon ratio values, 1970 to 1976 ....................................................... 93

C-3 Annual external coworker doses for plutonium workers, 1952 to 1970 ................................... 97

C-4 Annual external coworker doses for plutonium workers, 1971 to 2005 ................................... 99

C-5 Annual external coworker doses for uranium workers, 1952 to 2005.................................... 100

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

with ORAUT-OTIB-0052....................................................................................................... 101

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

with ORAUT-OTIB-0052....................................................................................................... 104

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

with ORAUT-OTIB-0052....................................................................................................... 104
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 7 of 105

LIST OF FIGURES

FIGURE TITLE

6-1 ORNL-style film badge ...........................................................................................................

PAGE

12

6-2 RFP multielement film badge ................................................................................................. 12

6-3 RFP interim TLD/film badge................................................................................................... 12

6-4 RFP Harshaw badge..............................................................................................................12

6-5 RFP Panasonic dosimeter...................................................................................................... 12

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

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

6-7 Estimated beta dose rate from uranium metal at various enrichment levels ........................... 52

6-8 Shallow dose rate from natural uranium slab.......................................................................... 53

A-1 Occupational Dose Report reviewed 6-4-02, page 1 .............................................................. 72

A-2 Occupational Dose Report reviewed 6-4-02, page 2 .............................................................. 73

A-3 Dosimetry History by Individual query report dated 3-10-03, page 1 ...................................... 74

A-4 Dosimetry History by Individual query report dated 3-10-03, page 2 ...................................... 75

A-5 Dosimetry History by Individual query report dated 3-10-03, page 3 ...................................... 76

A-6 Health Physics Yearly External Exposure Activity Run, 1953 to 1958 .................................... 77

A-7 Health Physics Yearly External Exposure Run, 1959 to 1963 ................................................ 78

A-8 Health Physics Yearly External Exposure Activity Run, 1964................................................. 79

A-9 External Dosimetry Detail, 1981 to 1983 ................................................................................ 80

A-10 Health Physics External Radiation Exposure Report, 1967 .................................................... 81

A-11 Radiation Dosimetry Individual Lifetime Report dated 6-4-02................................................. 82

A-12 Radiation Health Records System – TLD Data....................................................................... 83

A-13 Radiation Dosimetry Detail Report, Termination Report dated 9-17-96, page 1 ..................... 84

A-14 Radiation Dosimetry Detail Report, Termination Report dated 9-17-96, page 2 ..................... 85

A-15 Occupational Radiation Exposure Information, 1988.............................................................. 86
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 8 of 105

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

AP anterior-posterior

Bq becquerel

CHP Certified Health Physicist

Ci curie

cm centimeter

cpm counts per minute

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

dpm disintegrations per minute

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

Emax maximum energy

EU enriched uranium

ft foot

g gram

GM geometric mean

GSD geometric standard deviation

H*(d) ambient dose equivalent

HEU highly enriched uranium

HIS-20 Health Physics Information System (database)

Hp(d) personal dose equivalent

Hp,slab(d) personal dose equivalent (slab phantom)

HPS Health Physics Services (company)

hr hour

HSDB Health Sciences Database

ICRP International Commission on Radiological Protection

in. inch

IREP Interactive RadioEpidemiological Program

ISO isotropic

keV kiloelectron-volt, 1,000 electron-volts

kg kilogram

LANL Los Alamos National Laboratory

Lc critical level

LOD limit of detection

MeV megaelectron-volt, 1 million electron-volts
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 9 of 105

mg milligram

mm millimeter

mo month

mR milliroentgen

mrad millirad

mrem millirem

mSv millisievert

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-Office of Compensation Analysis and Support Claims Tracking System

NTA nuclear track emulsion, type A

ORNL Oak Ridge National Laboratory

ORAU Oak Ridge Associated Universities 

OSL Optically Stimulated Luminescence

OTIB ORAU Team technical information bulletin

PER program evaluation report

PNAD personal nuclear accident dosimeter

PNL Pacific Northwest Laboratory

POC probability of causation

ppb parts per billion

qtr quarter

R roentgen

RFETS Rocky Flats Environmental Technology Site

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

wk week

yr year

§ section or sections
Document No. ORAUT-TKBS-0011-6 Revision No. 02 PC-1 Effective Date: 10/20/2010 Page 10 of 105

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 historic background information and guidance to assist in the preparation of 

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2007b):

] guidelines established under 

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

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

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

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

conventional structures

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

reasons

 

1

 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 PC-1 Effective Date: 10/20/2010 Page 11 of 105

6.1.1

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 Section 6, Occupational External Dosimetry, of the Rocky Flats 

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. 

Purpose

6.1.2

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 the External Dose Reconstruction Implementation Guideline

(NIOSH 2007a). 

Scope

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 (2007a), 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 prior to 1970. Unmonitored and 

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

Employees Occupational Illness Compensation Program Act. Between 1967 and 1970, unmonitored 

and notional neutron doses should be replaced with external coworker doses. Reported NDRP and 

non-affected original neutron dose can be used during 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. 

6.2 EXTERNAL DOSIMETRY OVERVIEW

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

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 

and a 1-mm cadmium filter. For the plutonium areas, in 1960, a brass filter with half the filtration of 

the cadmium filter was added to cover half of the open window. 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 6-1). 
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 12 of 105

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 6-2). 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 open window. 

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

badge (Figure 6-3) until the introduction of the TLD neutron system (Baker 2002). 

Figure 6-1. ORNL-style film 

badge (including brass filter).

Figure 6-2. RFP multielement 

film badge.

Figure 6-3. RFP interim 

TLD/film badge. 

In 1971, a full TLD badge was introduced at RFP that 

used TLD chips manufactured by the Harshaw 

Chemical Company I(Figure 6-4). 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 (Figure 6-5). This badge 

contains two Panasonic dosimeters, one for measuring 

photon and beta dose and one for measuring neutron 

dose. The beta/photon dosimeter contains two TLD 

phosphors and a lead filter over one of the elements. 

The neutron dosimeter contains three neutronsensitive elements and one neutron-insensitive 

element under cadmium or tin filters. This badge

includes a PNAD (Baker 2002). 

Table 6-1 summarizes the history of dosimeter use at 

RFP. The implementation dates listed in the table and used throughout this document are not exact. 

In many cases, dosimeters were phased in over a period of 1 to 3 yr. Determining from an individual 

employee's dosimetry record which dosimeter was worn is not possible, which adds a degree of 

Figure 6-4. RFP Harshaw badge.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 13 of 105

uncertainty to dose reconstruction. Further research is necessary to identify exact dates for each 

dosimeter type.

The following sections discuss each of these dosimeter types in relation to each necessary dose 

reconstruction parameter. 

Figure 6-5. RFP Panasonic dosimeter.
Document No. ORAUT-TKBS-0011-6 Revision No. No. 02 Effective Date: 08/14/200

7 Page 14 of 105

Table 6-1. External dosimeter history. 

Year Holder

Beta/gamma

Neutron Extremity

Detector

Filtration

Deep Shallow Processor Detector Processor Holder Detector

1951a SS ORNL design Std. X-ray 1-mm Cd None LANL Track Plate LANL SS ORNL design Std. X-ray

1952 │ │ │ │ │ │ │ │ │

1953 │ │ │ │ │ │ │ │ │

1954 │ │ │ │ │ │ │ │ │

1955 │ │ │ │ │ │ │ │ │

1956 │ │ │ │ HPS │ HPS │ │

1957 │ │ │ │ │ NTA Film │ │ │

1958 │ │ │ │ RFETS │ RFETS │ │

1959 │ │ │ │ │ │ │ │ │

1960 │ │ 1/2 brassb │ │ │ │ 1/2 brassc │

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 │ │ │ │ │ │ │ │ │
Document No. ORAUT-TKBS-0011-6 Revision No. No. 02 Effective Date: 08/14/200

7 Page 15 of 105

Year Holder

Beta/gamma

Neutron Extremity

Detector

Filtration

Deep Shallow Processor Detector Processor Holder Detector

1991 (DOELAP) Panasonic UD-813AS11

1992 │ │ │ │ │ │ │ │ │

1993 │ │ │ │ │ │ │ │ │

1994 │ │ │ │ │ │ │ │ │

1995 │ │ │ │ │ │ │ │ │

1996 │ │ │ │ │ │ │ │ │

1997 │ │ │ │ │ │ │ │ │

1998 │ │ │ │ │ │ │ │ │

1999 │ │ │ │ │ │ │ (DOELAP) 

2000 │ │ │ │ │ │ │ │ │

2001 │ │ │ │ │ │ │ │ │

2002 │ │ │ │ │ │ │ │ │

2003 │ │ │ │ │ │ │ │ │

2004 │ │ │ │ │ │ │ │ │

2005 InLightTM OSLd │ │ Landauer CR-39 Landauer Luxel® OSL

a. Dates are approximate, overlap occurred during changeover (Baker 2002). 

b. Brass not used on beta open window (Baker 2002). 

c. Brass not used on beta open window, no brass on wrist side (Baker 2002).

d. OSL = Optically Stimulated Luminescence dosimeter.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 16 of 105

6.3 INTERPRETING THE EXTERNAL DOSIMETRY RECORD

When NIOSH requests an individual dosimetry record (file), the Rocky Flats Environmental 

Technology Site (RFETS) Radiological Health Department expends a significant amount of effort in 

reviewing and organizing the external dosimetry records. Both hard-copy and electronic files are 

reviewed. RFETS provides comments if discrepancies are found. If there are hard-copy results that 

are not in the electronic file, the electronic file is updated. If the electronic file includes a reading that 

is not indicated in the paper file, it is noted as a comment but left in place. The assumption favorable 

to claimants is to include discrepant data in the annual total, unless notes explain why the data should 

not be included [1]. 

External dosimetry results are reported as:

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

• Skin (shallow dose + neutron)

• Forearm (measured or estimated)

• 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 on a routine basis, but estimated the hand dose 

using the forearm dose measured by a wrist badge and the application of a hand-to-wrist ratio (see 

Section 6.10). 

6.3.1

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.

Dosimetry Records Systems 

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

• 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

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

dosimetry records (Author unknown, no date) [2].

Observed Data Discrepancies 
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 17 of 105

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.

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 Pos s ible Algorithm Is s ue

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 (RFP no date) 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 Pos s ible 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 from the archive at the Federal Center 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.

6.3.3

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.

Database Table-Specific Issues
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 18 of 105

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.

To migrate these data to the current electronic database, HIS-20, 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.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 19 of 105

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.

6.3.3.2.1 Dos e His tory Hard-Copy File Contents

The RFP Radiological Health organization reviews 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 are 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. A review of the rest of the external dosimetry file might indicate some 

detail of what went into the annual total. After 1975, this report provides 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 2003 and the result of review of these records by James M. Langsted 

[3]:

• 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 4, quarterly
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 20 of 105

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.

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

indicate the source of the dosimetry result [4]:

– 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 Radiological Health Records System (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

• 1977–2005, dosimeter exchange history
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 21 of 105

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

Table 6-2 provides detail for the interpretation of the values, zeros, and blanks encountered in the 

RFP reports detailed in the previous section.

Interpretation of Dosimetry Data

6.3.5

There are additional sources of information, which are known to exist [6], that contain detail that is not 

in the dose history file. These data might provide detail useful to refining dose estimates for some 

workers.

Additional Data Available

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 [7]. 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 Neutron Dose Reconstruction Project (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; therefore, 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).

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.
Document No. ORAUT-TKBS-0011-6 Revision No. No. 02 Effective Date: 08/14/2007 Page 22 of 105

Table 6-2. Interpretation of reported data [5]. 

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 6-11 & 6-18 and Section 

6.8.2.1.

Deep dose

Extremity dose

Skin dose (see Section 6.3 and 

6.6.1.2, last paragraph).

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

LDE–irrelevant to dose 

reconstruction.

included in 

DDE and SDE before 1977). 

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 not included in the 

DDE and SDE values. Otherwise, 

a zero indicates a monitored 

exposure reported as zero.

Blank indicates an 

unreported 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.

Health Physics External 

Exposure Activity Run

Yearly (Figure A-6)

Quarterly total of dose in mrem 

according to Tables 6-11 & 6-18 

and Section 6.8.2.1. 

Penetrating dose, Extremity dose 

(see Section 6.10), 

Skin dose (see Sections 6.3 & 

6.6.1.2, last paragraph). 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 6-11 & 6-18 and Section 

6.8.2.1. 

Zero indicates no measured dose 

under that filter

Blank indicates reading not 

available. For neutron 

badge (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.

Code Explanation

1 Gamma/beta dosimeter

2 Gamma/X-ray dosimeter

3 Neutron dosimeter

Dose column labels

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)
Document No. ORAUT-TKBS-0011-6 Revision No. No. 02 Effective Date: 08/14/2007 Page 23 of 105

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-8 and Section 

6.3.3.2)

Quarterly totals are mrem 

according to Tables 6-11 & 6-18 

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.

External Dosimetry (TLD) 

Detail

(Figure A-9 and Section 

6.3.3.2)

Individual dosimeter results are 

mrem according to Tables 6-11 &

6-18 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 6-11& 6-18 

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 

6-11 & 6-18.

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.
Document No. ORAUT-TKBS-0011-6 Revision No. No. 02 Effective Date: 08/14/2007 Page 24 of 105

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 6-11 &

6-18 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 

6-11 & 6-18.

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.

Occupational Radiation 

Exposure Information

(Figure A-15)

This report shows dose in rem Zeros indicate an annual total of 

zero based on both external 

dosimetry results of zero and 

unmonitored periods. 

according to Tables 6-11 & 6-18. 

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.

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.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 25 of 105

6.4 HISTORICAL ADMINISTRATIVE PRACTICES

6.4.1

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 991

Badged Population

2

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). 

received neutron dosimeters. The use of dosimetry expanded to other 

RFP production operations. 

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

external dosimetry. The results are shown in Figure 6-6. 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. 

0%

10%

20%

30%

40%

50%

60%

70%

80%

90%

100%

1951 1961 1971 1981 1991 2001

Year

Badged

EEOICPA Claims

Plant Population

Figure 6-6. Portion of plant population badged. 

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 

 

2 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. 02 Effective Date: 08/14/2007 Page 26 of 105

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

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 6-6a. 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 "diluted" by those monitored employees that did not 

work in the production areas.

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

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

6.4.2

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 

Badge Exchange Frequency
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 27 of 105

biweekly to semimonthly occurred. In later years, dosimetry was exchanged on semimonthly, 

monthly, and quarterly frequencies. In the 1990s, exchange frequencies went to monthly, quarterly, 

and semiannually. An option for annual exchanges was identified, but never used.

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 by James M. Langsted. 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 6-3 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.

6.4.3

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

know the facility in which the individual worked (no date) [8]. 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.

Field-Specific Calibration Factors

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

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

1992 [9]. 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 numbers to 

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

was reported as zero (RFETS 2001). This policy is consistent with the limits of detection (LODs)

reported elsewhere in this TBD.

Minimum Reported Dose
Document No. ORAUT-TKBS-0011-6 Revision No. No. 02 Effective Date: 08/14/2007 Page 28 of 105

Table 6-3. Conservatively determined default dosimeter exchange frequencies.

a,b

Year

Chemical 

operators

Metallurgical 

operators Maintenance 

workers

Support 

personnel

Analytical 

laboratory 

technicians

Site support 

personnel

Radiation 

control 

Pu U Pu U technicians D&D workers

1951 bw bw bw bw m m bw m bw m

1952 bw bw bw bw m m bw m bw m

1953 bw bw bw bw m m bw m bw m

1954 bw bw bw w m bw bw m bw bw

1955 bw bw bw w m bw bw m bw bw

1956 bw bw bw w m bw bw m bw bw

1957 bw bw bw w m bw bw m bw bw

1958 w w w w m bw bw m w bw

1959 w bw w w m bw bw m w bw

1960 w w w w m m bw m w m

1961 bw bw bw w m m w m bw m

1962 bw bw bw w m m w m bw m

1963 bw m bw bw m m q m bw m

1964 bw m bw m q m q q bw m

1965 m – m m q q q q m q

1966 m – m m m q q q m q

1967 bw – bw m m q q q bw q

1968 bw – bw m m m m m bw m

1969 bw – m m m m m m m m

1970 bw – m m m m m m m m

1971 bw – m m m m m m m m

1972 bw – m m m m m m m m

1973 bw – m m m m m m m m

1974 bw – m m m m m m m m

1975 bw – m m m m m m m m

1976 bw – m m m m m m m m

a. bw = Biweekly (every 2 wk), assumed because it is favorable to claimants over semimonthly (twice per month); m = monthly; q = quarterly; w = weekly.

b. Source: Study described in Section 6.4.2.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 29 of 105

6.4.5

Table 6-4 provides a summary of the calculations used to determine the recorded dose at RFP.

Recorded Dose Practices

Table 6-4. Summary of historical recorded dose practices.

a

Year Dosimeter measured quantities Compliance dose quantities

Two-element film (photon) + track plate (neutron)

1951–1956 Owdose = (Owdensity – Cddensity) × CFnet Ow

Cddose = Cddensity × CFCd

Ndose = neutron tracks × CFneutron

Pen = Cddose + 0.5 × Owdose + Ndose

Skin = Cddose + Owdose + Ndose

Two-element film + NTA film

1957–1959 Owdose = (Owdensity – Cddensity) × CFnet Ow

Cddose = Cddensity × CFCd

Ndose = neutron tracks × CFneutron

Pen = Cddose + 0.5 × Owdose + Ndose

Skin = Cddose + Owdose + Ndose

Two-element film (beta)

5/1953–10/1970 Owdose = (Owdensity – Cddensity) × CFnet Ow

Cddose = Cddensity × CFCd

(no neutron measured)

Pen = Cddose

Skin = Cddose + Owdose

Three-element film + NTA film

3/1960 (Building 71)

1/1963 (other Pu 

buildings)

2/1968 (Building 81 & 

Building 91)

–1962

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

Multiple-element film + NTA film

1963–1969 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

TLD + NTA film

1969–1970

1/1970 (Building 771)

4/1970 (other Pu 

buildings)

10/1970 (all others)b

–1970

Gdose = PTLD

Sdose = STLD

Ndose = neutron tracks × CFneutron

Pen = Gdose + Ndose 

Skin = Sdose + Ndose

TLD-700 + TLD-600/700

1971–1982 Gdose = PTLD

c

Sdose = STLD

Ndose = determined from albedo algorithm

Pen = Gdose + Ndose 

Skin = Sdose + Ndose 

if Skin < Pen, then Skin = Pen

Panasonic UD-802 + UD-809

1983–1989 Photon deep

Photon/Beta shallow

Neutron

Pen = photon deep + neutron

Skin = photon/beta shallow + neutron

1990–2004 Hs,gamma

Hd,gamma

Hs,beta

Ndose 

Hd = Hd,gamma + neutron

Hs = Hs,gamma + Hs,beta + neutron

a. Owdensity = open window (measured density); Owdose = open window (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-raysd; 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.

d. 1.14 or 1.17 or nonlinear factor.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 30 of 105

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

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 6-3 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. 

Number of Zero Readings

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. This would be the situation, for example, if an individual 

received a cumulative dose of 2,140 mrem in a given year at a facility that had a monthly monitoring 

frequency and where the maximum permissible exposure limit was 1,000 mrem/mo. The minimum 

number of months in which this dose could have been received is 3. Therefore, the maximum number 

of missed dose months would be 9, and the minimum would be zero because the dose could have

been received evenly throughout the year. The central estimated number of months should be the 

median, or 5; however, the upper bound would be 9 (NIOSH 2007a).

Quarterly or annual limits: 

• 1951–1967, 3 rem/qtr (Figure A-10)

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

• 1993–2005, 2 rem/yr (DOE 1992)

Table 6-5 divides these dose limits into exchange frequencies. Either the dosimetry records or the 

default values from Table 6-3 should be used to determine or estimate the exchange frequency and 

number of reported zeros. Using the methodology of NIOSH (2007a), it is possible to develop an 

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

6.5.2

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 [10]. 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 [11]. 

Discrepancies

Corrections were noted in the dose record when calculation or computer errors occurred [12]. 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.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 31 of 105

Table 6-5. 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.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 32 of 105

6.5.3

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. The 

assumption favorable to claimants is that the dosimeter was lost; dose should be assigned for that 

period using dosimetry data from before and after that period (dose reconstructors should consider 

the approach of Watson et al. 1994) or coworker data.

Missing Entry

6.5.4

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 2010). 

Exposure Geometry

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 2007a, 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 [13]. 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 6-6 

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 non-AP doses from 

the ROT geometry [14]. Table 6-7 lists these fractions, which are rounded.

Table 6-6. Exposure geometry calculation.

Major job category Hands-on work (time) Source geometry

Calculated dose received

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%

6.5.5

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 

Lead Aprons
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 33 of 105

Table 6-7. Exposure geometry defaults for major job 

categories.

Major job category

Default selected

AP ISO ROT

Chemical operators 50% 50%

Metallurgical operators 100%

Maintenance workers 100%

Support personnel 50% 50%

Analytical laboratory tech. 100%

Site support personnel 100%

Radiation control technicians 60% 40%

D&D workers 100%

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 

the dose to the torso [15]. 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 6-8 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 6-8. 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

Outside apron 1.9 1 1

Unprotected 

area

Under apron 1 1.2 1

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

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 

. It is 

appropriate to use these factors for photon doses measured with the film dosimeters.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 34 of 105

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 6-8 will be applied. 

6.5.6

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 

(RFETS 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 minimus levels. Recycled DU received 

from Fernald contained 2.8 ppb plutonium, 389 ppb neptunium, and 8,550 ppb technetium. A small 

quantity of recycled highly enriched uranium (HEU) received at RFP in 1955 contained 0.007 ppb 

plutonium, 2.5 ppb neptunium, and 9.12 ppb technetium. When contained within the uranium 

materials processed at RFP, these levels are insignificant in relation to external exposure (DOE 

2000).

Recycled Uranium

RFETS (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 (RFETS 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 (RFETS 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

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. 

Potential Elevated Background Subtraction
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 35 of 105

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 four retired dosimetry program managers. Each of these individuals 

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

recalled storage-board background as a problem. Most recalled that elevated storage background 

was not

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.

 significant and did not affect the dosimetry results [16].

From 1977 to February 2000, a plant-wide standard background was subtracted [17].

For dosimeters collected in March 2000 through 2003, badge storage-board background dosimeter 

results were used. The background dosimeter results were averaged over a five-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 (one sigma) times the previous (1977 to February 2000) standard 

background [18]. This information indicates that the background is in excess of that identified in the 

Technical Basis Document for Rocky Flats Plant Occupational Environmental Dose (ORAUT 2006b, 

Table 4-3). As indicated above, this dose was subtracted as dosimeter background and indicates an 

elevated ambient level of external radiation (ORAUT 2003a). 

6.5.8

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 be read unless 

circumstances warrant. A zero was then recorded in the workers record. Therefore, all 

reported zeros for film badges between 1969 and 1970 are suspect and should not be used in 

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

Badge Reading Policy in 1969-1970

6.6 PHOTON DOSE

6.6.1

The NIOSH Interactive RadioEpidemiology Program (IREP) software for calculating the POC (NIOSH 

2007a) contains three photon energy bands:

Energy Groups

• Below 30 keV

• 30 to 250 keV

• Above 250 keV 

Separation of the dose from each energy band is required.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 36 of 105

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, MicroShield 5.03 (Grove Engineering 1998) 

was used. With the use of the MicroShield decay feature, radionuclide source concentrations (DOE 

1980) for weapons-grade plutonium, enriched uranium, 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. Depleted and enriched uranium were routinely handled in 

the open with no shielding. Plutonium was almost exclusively handled in gloveboxes that provided

shielding from the materials. The MicroShield 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 6-9 presents these results.

Table 6-9. Photon energy distributions.

a

Shield

Energy

(keV)

Fresh 

Pu

10-yr 

Pu

30-yr 

Pu

Fresh 

EU

10-yr 

EU

30-yr 

EU

Fresh 

DU

10-yr 

DU

30-yr 

DU

None

<30

Not applicable

0% 0% 0% 0% 0% 0%

30–250 100% 99% 98% 100% 3% 3%

>250 0% 1% 2% 0% 97% 97%

1/16 

inch 

steel

<30 0% 0% 0%

30–250 100% 85% 88% Not applicable

>250 0% 15% 11%

a. Source: EU = enriched uranium.

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 [19]. 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 

Table 6-9. 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 enriched uranium 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 enriched uranium, 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 6-10.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 37 of 105

Table 6-10. Default photon energy distributions. 

Energy

(keV) Plutonium EU DU

<30 25% 0% 0%

30–250 75% 100% 50%

>250 0% 0% 50%

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) [20].

6.6.1.2.1 Pre -1960

The two-element film dosimeter used at RFP prior to 1960 had both open-window (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). 

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: 

electrons>15 keV or photon<30 keV = [Skin − Pen]/0.5 = OW (6-1)

photon(30 keV-250 keV ) + photon(>250 keV ) = Pen = CD + 50%*OW (6-2)

The equations above should be used with the facility radiation characteristics listed for Rocky Flats 

facilities in Tables 6-10 for plutonium, depleted uranium, or enriched uranium facilities to assign 

measured dose in the appropriate electron and photon ranges and proportions. The use of these 

equations may 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 1960–1970

The three-element film dosimeter used at RFP from 1960 to 1969 also had an Open Window (OW), a 

cadmium (CD) filter, and an additional brass (BR) filter providing half the filtration of the CD. The 

brass filter was added to more accurately measure the 60-keV photons. This dosimeter was phased 

out during the year 1970.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 38 of 105

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:

electron>15 keV or photon<30 keV = [Skin − Pen]/0.65 =OW (6-3)

photon30 keV–250 keV + photon>250 keV = Skin – (electron>15 keV or photon<30 keV ) = BR + CD (6-4)

The equations above should be used with the facility radiation characteristics listed for Rocky Flats 

facilities in Tables 6-10 for plutonium, depleted uranium, or enriched uranium 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–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. Since doses in the era of TLD dosimetry at RFP are felt to provide 

accurate measurements of skin and deep doses, the following relationships should be used for this 

period:

electrons>15 keV or photon<30 keV = Skin − Pen (6-5)

photon30 keV–250 keV + photon>250 keV = Pen (6-6)

The equations above should be used with the facility radiation characteristics listed for Rocky Flats 

facilities in Tables 6-10 for plutonium, depleted uranium, or enriched uranium facilities to assign 

measured dose in the appropriate electron and photon ranges and proportions.

6.6.2

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

Calibration Factor

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. 

The LANL dosimetry results were calibrated in roentgens (ORAUT 2009, Section 6E.9).

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) 
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 39 of 105

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 6-11 

summarizes dose units to use for organ dose conversion factors.

Table 6-11. Photon dose units for use 

with organ dose conversion factorsa

.

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 2007a. 

Plutonium-specific photon dose conversion factors are provided in Table 4.1a of NIOSH 2007a and 

should be applied for plutonium exposures at RFP.

6.6.3

Section 2.1.2 of NIOSH (2007a) recommends the use of the LOD/2 method for determining missed 

dose.

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 open window and an 

area covered with 1 mm of silver, tin, or cadmium (Alvarez et al. 2003). A PNL study of this twoelement 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), 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 [21].

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 6-12 lists photon LODs for the RFP dosimeters.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 40 of 105

Table 6-12. 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 the External Dose Reconstruction Implementation Guideline (NIOSH 

(2007a). For the period from 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 

(2007a) discuss this estimate. In this case, the estimate is assumed to have a lognormal distribution

(NIOSH 2007a, 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.

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 2005b) discusses the response of this dosimeter. 

These documents address the significant over-response of film to low photon energies. The 

dosimeter (open window) was calibrated with low-energy photons. To correct for this over-response, 

a portion of the open-window dose was added to the deep dose measured under the 1-mm cadmium 

filter. There is evidence (Falk, no date) [22] 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 
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 41 of 105

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

6.6.4.1 Angular Dependence

Geometry

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 over-response.

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 [23].

Quantitative information is available for the RFP Panasonic dosimeter (RFETS 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.

6.6.4.2 Exposure Geometry

A 100% anterior-to-posterior exposure geometry for all external doses at Rocky Flats should be 

applied, in accordance with Section 6.5.4.

6.6.5

The External Dose Reconstruction Guideline (NIOSH 2007a) describes methods for quantification of 

laboratory uncertainty associated with reading film and TLDs. These methods provide a statistical 

treatment of the variability associated with reading dosimeters in the laboratory.

Uncertainty

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 (NRC 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 (2007a) 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 (NRC 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 
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 42 of 105

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 6-13), and it affects the estimate of 

the upper 95% confidence dose above 27 mR. Table 6-13 lists uncertainties for photon film dose.

Table 6-13. 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

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 Loos e-Chip Thermolumines cent Dos imeters

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 * mean were removed from use. Assuming that the chip response had a normal distribution

such 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) ÷ 
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 43 of 105

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 (2007a), and assuming the critical level (Lc) is the LOD 

estimated in Section 6.3.1 of this TBD, Table 6-14 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. 

Table 6-14. 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

6.6.5.2.2 Panas onic Thermolumines cent Dos imeter

Table 6-15 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 (2007a). Uncertainty is 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 (RFETS 2001), and the relative standard deviation at high readings 

is the standard deviation of the DOELAP performance test results (RFETS 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 [24].

Table 6-15. Uncertainty for DOELAP-accredited TLD photon dose. 

Dose (mrem)

Upper 95% confidence dose (mrem)

Panasonic dosimeter DOELAP-accredited Panasonic dosimeter

1983–1989 1990–1998 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.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 44 of 105

A default value of 1.23 for 1983-1998 and 1.18 for 1999-2004 can be used for the photon TLD dose 

uncertainty multiplier.

6.7 NEUTRON DOSE

6.7.1

The measured neutron dose must be divided into energy groups consistent with the dose conversion 

factors provided in Appendix B of NIOSH (2007a). These energy groups and the associated radiation 

weighting factors wR from International Commission on Radiological Protection (ICRP) Publication 60 

(ICRP 1991) are:

Energy Groups

• <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)

• >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 

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 6-16 

lists this information.

Table 6-16. Neutron dose measurements divided into energy groups. 

Location

Dose rate

(rem/hr)

Avg. 

energy

(MeV)

Portion of dose from neutron energy range

< 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 0.006 0.006 0.437 0.552 0.000

Building 707 low dose pit 2.88E-04 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 0.035 0.017 0.687 0.261

Standard deviation 0.031 0.010 0.117 0.130
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 45 of 105

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 6-16. Table 6-17 lists the default values selected from Table 

6-16 for dose reconstruction [25].

The doses and fractions discussed above are based on quality factors published in NCRP (1971). 

NIOSH (2007a) indicates the use of radiation weighting factors 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 radiation weighting factor 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 6-17 

lists the multipliers that were determined for each neutron energy interval. The fraction of the dose 

Table 6-17. Default neutron energy distribution. 

Neutron energy 

intervals

Fraction of dose 

(NCRP 38)

Dose multiplier 

(ICRP 60) Dose multiplier

a

<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. 

using NCRP (1971) quality factors and the dose multiplier using ICRP (1991) radiation weighting 

factors were combined to determine a dose multiplier (column 4 of Table 6-17). 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.

6.7.2

6.7.2.1 Dosimeter-Specific Quality Factor Conversion

Calibration Factor

The correction factors to convert from NCRP (1971) quality factors used in the neutron spectra 

measurements and the ICRP (1991) radiation weighting factors are discussed in Section 6.7.1.2 and 

listed in Table 6-17. Conversion to organ dose is accomplished using the factors provided in 

Appendix A of NIOSH 2007a. 

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 2009, Section 6E.9).

Neutron film was initially calibrated with an apparently unmoderated polonium-beryllium (PoBe) 

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). 
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 46 of 105

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 1968). 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 6-18 summarizes the dose units to use for organ dose conversion factors.

Table 6-18. Neutron dose units for use 

with organ dose conversion factors.

Period Unit

1951–1983 H*(10)

1983–2005 Hp,slab(10)

6.7.3

6.7.3.1 Limit of Detection

Missed Dose

LANL processed neutron track plates for RFP from 1951 to 1956. The performance of this system is 

documented in Section 6E.7 of ORAUT (2009). The minimum detectable dose is identified as 

<50 mrem.

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 (ORISE 2005) assembled a 

processing history that is summarized in Table 6-19. 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 6-19. Neutron film track counting detail.

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 –

Mann and Boss (1963) determined that a typical background film for 2 wk 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. 
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 47 of 105

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 (RFETS 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 6-20 includes this value.

Table 6-20. 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

6.7.3.2 Number of Zero Readings

Section 6.5.1 of this TBD discusses the number of zero readings.

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 
Document No. ORAUT-TKBS-0011-6 Revision No. 02 PC-1 Effective Date: 10/20/2010 Page 48 of 105

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 (ORISE 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

The total neutron dose needed for dose reconstruction consists of three quantities that are listed in 

the summary and detail presentations. These quantities are:

 – Further details by exchange period in a given year about the original 

neutron dose, calculation of NDRP dose, the building location, and penetrating gamma dose

• 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 ORISE (2005) contains a comprehensive 

description of the notional dose calculation. A blank in this field indicates that either 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.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 PC-1 Effective Date: 10/20/2010 Page 49 of 105

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 the 

External Dose Reconstruction Implementation Guideline (NIOSH 2002).

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.

• 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. Coworker data should 

be used in reconstructing unmonitored neutron dose during this period.

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 should apply the 

approach from OCAS-IG-001 (NIOSH 2002). 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.

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 6-21 and 6-22 list the values for the periods from 1970 to 1976 

and from 1977 to 2000, respectively.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 50 of 105

Table 6-21. RFP 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/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 1.32b 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. Data for 1973 were not available. This value is the greater of the ratios for 1972 and 1974.

Table 6-22. ORAU Team-developed neutron-to-gamma ratios. 

6.7.4

6.7.4.1 Angular Dependence

Geometry

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.

6.7.4.2 Exposure Geometry

A 100% anterior-to-posterior exposure geometry for all external doses at Rocky Flats should be 

applied, in accordance with Section 6.5.4.

6.7.5

6.7.5.1 Film

Uncertainty

The NDRP has evaluated film uncertainty. For this reason, uncertainty of film neutron dosimeters is 

not discussed in this technical basis document. 
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 51 of 105

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 (2007a) is used.

The initial Panasonic TLD algorithm was evaluated during development (RFP, no date). 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 (RFETS 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 6-23 lists the uncertainties for these dosimetry systems.

Table 6-23. Uncertainty for TLD neutron dose measurements.

Dose 

(mrem)

Upper 95% confidence dose (mrem)

Harshaw TLD 

dosimeter

Panasonic 

dosimeter DOELAP-accredited Panasonic dosimeter

1971–1982 1983–1990 1991–1992 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.

A default value of 1.2 for 1971-1982, 1.21 for 1983-1990, 1.14 for 1991-1992, and 1.18 for 1993-2004 

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 
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 52 of 105

assumed that the skin dose reported in RFP dosimetry records for uranium workers results from 

electrons with energies greater than 15 keV. 

Figure 6-7 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 

(Emax) beta particles from 234mPa, the most energetic contributor to the beta exposure.

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 [26].

6.8.1

6.8.1.1 Exposure Spectra

Energy Groups

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 d) and can be conservatively assumed to be present at these levels. 

Figure 6-7. Estimated beta dose rate from uranium metal at various enrichment 

levels (DOE 2001).

Figure 6-8 shows the relative dose rate in relation to energy. DU would be similar to the natural 

uranium used for this experiment.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 53 of 105

Figure 6-8. Shallow dose rate from natural uranium slab (DOE 2001). 

6.8.1.2 Reported Dose to Energy Groups

NIOSH (2007a) 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 Interpretation of Dosimetry Data For Assignment of Shallow 

Dose (ORAUT 2005b).

6.8.2

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

Calibration Factor

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

6.8.3

per hour. The radiation weighting factor 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. 

6.8.3.1 Limit of Detection

Missed Dose

Beta dosimetry at RFP used open-window film calibrated to a uranium slab. ORAUT (2007a, 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 (2007a, Section 6.5.2) states 

that the minimum detectable dose would have been similar to that for photons. Wilson et al. (1990) 
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 54 of 105

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 6-24.

Table 6-24. 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 open-window (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 (RFETS 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 6-24.

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 [27]. There is 

therefore no unmonitored energy range for which a correction factor is appropriate.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 55 of 105

6.8.4

6.8.4.1 Angular Dependence

Geometry

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 (RFETS 2001, Section 04.06.2). However, based on the averaging effect cited in 

DOE (2001), no angular correction factor is proposed.

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

The method in NIOSH (2007a) 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 (NRC 1989).

Uncertainty

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 the External Dose Reconstruction 

Implementation Guideline (NIOSH 2007a) 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 (NRC 1989). This TBD analysis used a similar uncertainty estimation 

methodology and developed a spreadsheet that matched the illustration given in NRC (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 

6-25 lists these upper 95% confidence doses.

Table 6-25. 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
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 56 of 105

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 Loos e-Chip Thermolumines cent Dos imeters

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 (2007a), and assuming that the critical level (Lc) is the beta LOD estimated in 

Section 6.8.3.1 of this TBD, Table 6-26 lists the upper 95% confidence dose.

6.8.5.2.2 Panas onic TLD Dos imeter

Table 6-27 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 (2007a). 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 (RFETS 2001), and the relative 

standard deviation at high readings is the standard deviation of DOELAP performance test results 

Table 6-26. Uncertainty for loosechip 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

Table 6-27. Uncertainty for DOELAP-accredited TLD beta dose.

Dose (mrem)

Upper 95% confidence dose (mrem)

Panasonic dosimeter DOELAP-accredited Panasonic dosimeter

1983–1989 1990–1998 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
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 57 of 105

2,000 2,243 2,243 2,231

a. 1.12 multiplier for any dose greater than 2 mrem.

(RFETS 2001; Stanford 1990). The reasonable worst-case value (high-energy photons + neutrons 

mixture from RFETS 2001, Table 11) was selected to provide a result that is favorable to claimants.

6.8.6

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 (2007a).

Skin Contamination

DU 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.

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 upon 

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 6-28 lists the other isotopes.

Table 6-28. DU mixtures (DOE 1980).

Isotope

Mixture

Ci/g (mix)

Alpha Beta nCi/g (mix)

Th-231 4.90E-09

Th-234 3.40E-07 340

U-234 3.70E-08 37

U-235 4.90E-09 4.9

U-238 3.40E-07 340

Total 3.82E-07 3.45E-07 726.8

It is favorable to claimants to assume that the DU is 1 yr old. This allows for ingrowth of progeny to 

achieve secular equilibrium. A decay calculation using MicroShield 5.03 (Grove Engineering 1998) 

was performed. Table 6-29 lists the full set of decay isotopes. 

Table 6-29. One-year-old DU.

Nuclide

per gram of DU

curies becquerels

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

Nuclide

per gram of DU

curies becquerels

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

Nuclide

per gram of DU

curies becquerels

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

Po-215 1.28E-15 4.75E-05
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 58 of 105

Nuclide

per gram of DU

curies becquerels

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

Nuclide

per gram of DU

curies becquerels

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

Nuclide

per gram of DU

curies becquerels

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 2007a) or other 

appropriate means.

6.9 UNMONITORED INDIVIDUALS

6.9.1

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. 

In Production Areas

6.9.2

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. 

Outside Production Areas

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 2007b)].

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 open-window (thin Mylar) 

version.

In 1991, RFP switched to a Panasonic model UD-813AS11 (custom design) dosimeter in a plastic 

wrist holder (RFETS 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 

open window for beta and low-energy photon dose measurements. The dosimeter, which has 

undergone DOELAP performance testing, is documented in RFETS (2000b). 

RFP never used finger rings on a routine basis but estimated the hand dose using the forearm dose 

measured by the wrist badge and the application of a hand-to-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 

RFETS (2000b) indicates a ratio of 3 was implemented in approximately 1992 as a conservative 

estimate based on the results of several studies.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 59 of 105

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; RFETS 2000b).

Additional information on these dosimeters will be required for dose reconstruction for shallow dose to 

the extremity, if necessary.

6.11 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in the preceding text, bracketed callouts have been inserted to indicate 

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

reconstruction. These callouts are listed here again in the Attributions and Annotations section of the 

document, with information provided to identify the source and justification for each associated item. 

Conventional references are provided in the next section of this document, linking data, quotations, 

and other information to documents available for review on the ORAU Team servers.

Much of the information in this TBD was authored by James M. Langsted, Certified Health Physicist 

(CHP), and some is based on his recollections of his operations experience and administration of the 

programs in the radiation dosimetry programs at RFP.

[1] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Discussion in June 2005 with Ken Savitz, RFP dosimetry database and records professional, 

indicated that dosimetry results that are posted in the paper-copy record but not included in 

the HIS-20 database are noted on the dosimetry data review sheet. It is favorable to claimants 

to use this data because in this case the worker would be credited for this dose, rather than 

not considering that dose. 

[2] Author unknown (no date) was transmitted by e-mail. It has been submitted to the ORAU 

Team Site Research Database (SRDB) and is accessible as a formal reference. 

[3] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

During the development of this document, Mr. Langsted reviewed many NIOSH claim files and 

interpreted the dosimetry record contained in those files based on his experience with the 

dosimetry programs and recordkeeping systems at RFP.

[4] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Mr. Langsted was involved in the initial implementation of the Panasonic dosimetry system in 

1989. During the transition period it was necessary to distinguish the source of the dosimetry 

data (the new Panasonic system or the old loose-chip Harshaw system). This code scheme 

was identified as a solution and implemented by the programmers during the database 

modification necessary to accept the Panasonic dosimetry system data.

[5] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

This table was developed by Mr. Langsted based on his review of the reports in various 

NIOSH claim files and his understanding of the dosimetry recordkeeping processes used at 

RFP. In many of these cases, Mr. Langsted does not have direct experience during the 

periods when those reports were generated but is making assumptions based on the pattern 

of data shown in these worker reports.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 60 of 105

[6] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Mr. Langsted is aware of each of these additional sources of data as documented in each of 

the explanatory paragraphs. 

[7] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Mr. Langsted has discussed the job history card with the previous manager of Radiological 

Health at RFP, and the manager of the RFP NDRP. Both have verified the existence and 

content of these data. 

[8] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Mr. Langsted has discussed the processing of film with a previous professional health 

physicist in the RFP dosimetry department during part of the film-processing period. This 

individual verified that the film calibration factors (beta or soft gamma) used for processing a 

specific badge were selected based on the building to which the worker was assigned.

[9] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Mr. Langsted has reviewed many external dosimetry reports and NIOSH claim files. In all of 

these reports it is clear that the doses were reported down to zero with no reporting threshold. 

Doses ranging from 1 mrem upward have been observed.

[10] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Mr. Langsted has reviewed many external dosimetry reports and NIOSH claim files. In all of 

these reports it is clear that unlabeled doses are in millirem. If they were in rem, there would 

be many exposures in excess of the regulatory limits (2 rem/qtr or later 5 rem/yr). If this were 

the case, there would be evidence of concern (and investigation) recorded on at least some of 

the reports.

[11] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Several times in 2006, Mr. Langsted has talked with the manager of the RFP dosimetry 

organization from approximately 1958 to 1969. This individual corroborated that 

overexposures would have been noted on the reports and in the hard-copy health physics file.

[12] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Mr. Langsted has observed many indications of doses corrected in the health physics (written) 

files. There was apparently a computer error that resulted in incorrect dose reported on the 

printed record. This error was discovered and corrected in both the computer file (as noted on 

the printed record) and on the printed record itself.

[13] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Mr. Langsted has made estimates based on his observation of operations in the production 

buildings during his experience at RFP as an Operational Health Physicist.

[14] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Mr. Langsted’s observation of the production areas indicates that Building 771 had 

considerable process piping in the overhead (above the worker’s head) while Buildings 

776/777 and Building 707 (metal production) had considerably less. As discussed in the text, 

the ISO geometry accounts for overhead exposure while the ROT geometry does not.

[15] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

In 2005, Mr. Langsted interviewed a previous Health Physics manager and discussed the use 

of lead aprons at RFP. This individual recalled the use of lead aprons as discussed in the text.

[16] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 61 of 105

In 2005, Mr. Langsted interviewed four Health Physics managers. Each of these individuals 

was asked specifically about this issue and replied in the negative.

[17] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Mr. Langsted has reviewed RFP dosimetry program memoranda available on the ORAU Team 

servers and found indication of the start and end dates of this practice.

[18] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

In 2005, Mr. Langsted interviewed Ken Savitz from the RFP radiological health organization 

and determined that a 5-qtr rolling average was calculated for badge board background. This 

information was obtained from the external dosimetry group at RFP and analyzed in a 

spreadsheet to determine the variability indicated in the text. 

[19] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

This assumption was developed in discussion with Tim Taulbee, a NIOSH health physicist. 

This value is based on assumptions as discussed in the text and review of other ORAU Team 

TBDs for plutonium facilities. The value was assumed at the direction of Mr. Taulbee after his 

discussion with other NIOSH health physicists.

[20] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

This analysis was provided by Tim Taulbee, NIOSH health physicist. Mr. Taulbee performed 

the analysis using the skin dose as reported by the RFP dosimetry program (Table 6-4).

[21] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

The Hanford dosimeter used similar dosimeter elements (Harshaw chips). Although the 

process equipment was somewhat different, the physics of the dosimetry process is the same 

and the LOD was assumed by Mr. Langsted to be similar.

[22] Falk (no date) was captured from the personal files of Roger B. Falk and has been 

documented in the ORAU Team SRDB as a formal reference.

[23] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

No measurement of the TLD badge exposure geometry has been found. As the badge is 

rotated in relation to the exposure source, the design of the badge is such that the TLD chip is 

exposed with less of the shield covering the chip. This would exhibit similar edge-effect 

response as would the film badge.

[24] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

The same badge system, including dosimeter type, case, dosimeter reader, and calibration 

system, was used. Although differing procedures were used for the program, Mr. Langsted 

believes it is reasonable to assume that the reading variability would be similar. 

[25] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Mr. Langsted could identify no discernable pattern in the data presented in Table 6-16. 

Therefore, he determined that using the mean values would provide the most representative 

neutron distribution for the neutron fields presented in the cited study. 

[26] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Mr. Langsted worked as an Operational Health Physicist at RFP in Building 444. At that time, 

uranium foundry workers wore leather gloves while working with the material. It was 

necessary to dispose of the gloves periodically when the dose rate measured on them

exceeded a certain level. This limited the hand doses from the uranium and uranium progeny 

from the gloves.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 62 of 105

[27] Langsted, James M., CHP. MH Chew & Associates. Principal Health Physicist. July 2006.

Dosimeters that measure skin dose must have a thin covering over the dosimeter to measure 

accurately the low-energy radiations that result in a component of the skin dose. This fact has 

been recognized and incorporated into the design of all dosimeters at RFP. This was 

implemented as open-window film (covered only with paper of the film packet) or the thinly 

covered skin TLD crystals. The very-low-energy beta particles that do not penetrate these 

coverings similarly cannot penetrate the protective epidermal layer of the skin and do not 

result in biologically significant skin dose. 
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 63 of 105

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Mann, J. R., and M. R. Boss, 1963, Neutron Dosimetry Problems in a Plutonium Processing Plant, 

RFP-305, 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.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 PC-1 Effective Date: 10/07/2010 Page 65 of 105

NIOSH (National Institute for Occupational Safety and Health), 2010, Best Estimate External Dose 

Reconstruction for Glovebox Workers, DCAS-TIB-0010, Rev. 03, Cincinnati, Ohio, June 18. 

NIOSH (National Institute for Occupational Safety and Health), 2007a, External Dose Reconstruction 

Implementation Guideline, OCAS-IG-001, Rev. 3, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, November 21. 

NIOSH (National Institute for Occupational Safety and Health), 2007b, Radiation Exposures Covered 

for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, OCAS-IG-003, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, November 5. [SRDB Ref ID: 35987]

NRC (National Research Council), 1989, Film Badge Dosimetry in Atmospheric Nuclear Tests, 

National Academy of Sciences, National Academy Press, Washington, D.C. [SRDB Ref ID: 

15199]

ORAUT (Oak Ridge Associated Universities Team), 2007a, Oak Ridge National Laboratory – 

Occupational External Dose, ORAUT-TKBS-0012-6, Rev. 01, Oak Ridge, Tennessee, 

September 10. 

ORAUT (Oak Ridge Associated Universities Team), 2008, Use of Coworker Dosimetry Data for 

External Dose Assignment, ORAUT-OTIB-0020, Rev. 02, Oak Ridge Tennessee, December 4. 

ORAUT (Oak Ridge Associated Universities Team), 2005a, Interpretation of Dosimetry Data for 

Assignment of Shallow Dose, ORAUT-OTIB-0017, Rev. 01, Oak Ridge, Tennessee, 

October 11.

ORAUT (Oak Ridge Associated Universities Team), 2009, Los Alamos National Laboratory – 

Occupational External Dose, ORAUT-TKBS-0010-6, Rev. 02, Oak Ridge, Tennessee, 

November 23. 

ORAUT (Oak Ridge Associated Universities Team), 2005b, Savannah River Site, ORAUT-TKBS0003, Rev. 03, Oak Ridge, Tennessee, April 5.

ORAUT (Oak Ridge Associated Universities Team), 2006a, A Standard Complex-Wide Methodology 

for Overestimating External Doses Measured with Thermoluminescent Dosimeters, ORAUTOTIB-0008, Rev. 01, Oak Ridge, Tennessee, May 12.

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. 

ORAUT (Oak Ridge Associated Universities Team), 2007c, Rocky Flats Plant -- Occupational 

Environmental Dose, ORAUT-TKBS-0011-4, Rev. 02, Oak Ridge, Tennessee, April 23.

ORAUT (Oak Ridge Associated Universities Team), 2007d, Parameters to Consider When 

Processing Claims for Construction Trade Workers, ORAUT-OTIB-0052, Rev. 00 PC-1, Oak 

Ridge, Tennessee, January 16. 

ORISE (Oak Ridge Institute of Science and Education), 2005, Technical Basis Document for the 

Neutron Dose Reconstruction Project, ORISE 05-0199, Oak Ridge, Tennessee, February 7.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 66 of 105

Passmore, C. N., 1992, RFP Lead Apron Field Study (TLD Inside vs. Outside Lead Apron), Rocky 

Flats Plant Site, Golden, Colorado, March 21. [SRDB Ref ID: 22289]

Putzier, E. A., 1982, The Past 30 Years at Rocky Flats Plant, Rockwell International Energy Systems 

Group, Rocky Flats Plant, Golden, Colorado, November. [SRDB Ref ID: 4632]

RFETS (Rocky Flats Environmental Technology Site), 2000a, Report on the Flow of Recycled 

Uranium at the Department of Energy's Rocky Flats Plant, 1953 - 1993, Golden, Colorado, 

June 30. [SRDB Ref ID: 16227]

RFETS (Rocky Flats Environmental Technology Site), 2000b, Rocky Flats Environmental Technology 

Site External Dosimetry Technical Basis Document - Extremity Dosimetry, ED-TBM-05, Rev. 

02, September 25. [SRDB Ref ID: 8797, p. 59]

RFETS (Rocky Flats Environmental Technology Site), 2001, External Dosimetry Technical Basis 

Document - Whole Body Dosimetry, ED-TBM-04, Rev. 2, September 25. [SRDB Ref ID: 

8797, p. 34]

RFP (Rocky Flats Plant), no date, Algorithm for Panasonic 802 and 809 Dosimeters for the Rocky 

Flats Dosimetry Badge. [SRDB Ref ID: 4627]

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]

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, Department of Preventive Medicine and Biometrics, University 

of Colorado Health Sciences Center, and Colorado Department of Public Health and 

Environment. [SRDB Ref ID: 8060]

Savitz, K., 2003, Explanation of the Electronic Data, August 5. [SRDB Ref ID: 24757]

Stanford, N., 1990, Whole Body TLD Dose Algorithm for Panasonic 802/809 TLD Pair, Rocky Flats 

Plant External Dosimetry Program, September 30. [SRDB Ref ID: 4641]

Watson, J. E., Jr., J. L. Wood, W. G. Tankersley, and C. M. West, 1994, "Estimation of Radiation 

Doses for Workers without Monitoring data for Retrospective Epidemiologic Studies," Health 

Physics, volume 67, number 4, pp. 402–405. [SRDB Ref ID: 8601]

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

U.S. Atomic Energy Commission (AEC)

Original agency established for nuclear weapons and power production; a predecessor to the 

U.S. Department of Energy.

beta dose

A designation (i.e., beta) on some external dose records referring to the dose from less

energetic beta, X-ray, and/or gamma radiation (see open window and shallow dose). 

beta radiation

Radiation consisting of charged particles of very small mass (i.e., the electron) emitted 

spontaneously from the nuclei of certain radioactive elements. Most (if not all) of the direct 

fission products emit beta radiation. Physically, the beta particle is identical to an electron 

moving at high velocity.

deep absorbed dose

The absorbed dose at the depth of 1.0 cm in a material of specified geometry and 

composition.

deep dose equivalent (DDE, Hd)

The dose equivalent at the respective depth of 1.0 cm in tissue.

dose equivalent (H)

The product of the absorbed dose D, the quality factor Q, and any other modifying factors. 

The special unit is the rem. When D is expressed in grays, H is in sievert. 

(1 sievert = 100 rem.)

dosimeter

A device used to measure the quantity of radiation received. A holder with radiation-absorbing 

elements (filters) and an insert with radiation-sensitive elements packaged to provide a record 

of absorbed dose or dose equivalent received by an individual. (See albedo dosimeter, film 

dosimeter, neutron film dosimeter, thermoluminescent dosimeter.)

dosimetry

The science of assessing absorbed dose, dose equivalent, effective dose equivalent, etc., 

from external or internal sources of radiation. 

dosimetry system

A system used to assess dose equivalent from external radiation to the whole body, skin, or 

extremities. This includes the fabrication, assignment, and processing of dosimeters as well 

as interpretation and documentation of the results.

exchange period (frequency)

Period (weekly, semimonthly, biweekly, monthly, quarterly, etc.) for routine exchange of 

dosimeters.

exposure

As used in the technical sense, exposure refers to a measure expressed in roentgens of the 

ionization produced by photons (i.e., gamma and X-rays) in air. 
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 68 of 105

extremity

That portion of the arm extending from and including the elbow through the fingertips, and that 

portion of the leg extending from and including the knee and patella through the tips of the 

toes.

field calibration

Dosimeter calibration based on radiation types, intensities, and energies in the work 

environment.

film

In general, a film packet that contains one or more pieces of film in a light-tight wrapping. 

When developed, the film has an image caused by radiation that can be measured using an 

optical densitometer.

film density

See optical density.

film dosimeter

A small packet of film in a holder that attaches to a wearer.

fission

The splitting of a heavy atomic nucleus, which is accompanied by the release of energy.

fissionable

Material capable of undergoing fission.

gamma rays

Electromagnetic radiation (photons) originating in atomic nuclei and accompanying many 

nuclear reactions (e.g., fission, radioactive decay, and neutron capture). Physically, gamma 

rays are identical to X-rays of high energy, the only essential difference being that X-rays do 

not originate in the nucleus. 

ionizing radiation

Electromagnetic or particulate radiation capable of producing charged particles through 

interactions with matter.

isotope

Elements having the same atomic number but different atomic weights; identical chemically 

but having different physical and nuclear properties.

neutron

A basic particle that is electrically neutral weighing nearly the same as the hydrogen atom.

neutron film dosimeter

A film dosimeter that contains a neutron track emulsion, type A film packet.

nuclear track emulsion, type A (NTA)

A film that is sensitive to fast neutrons. The developed image has tracks caused by neutrons 

that can be seen by using an appropriate imaging capability such as oil immersion and a 1000-

power microscope or a projection capability.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 69 of 105

open window (OW)

Designation on film dosimeter reports that implies the use of little (i.e., only security credential) 

shielding. Commonly used to label the film response corresponding to the open-window area. 

operating area

Designation of major onsite operational work areas.

optical density

The quantitative measurement of photographic blackening; density defined as D = log10 (Io/I).

personal dose equivalent, Hp(d)

Represents the dose equivalent in soft tissue below a specified point on the body at an 

appropriate depth d. The depths selected for personnel dosimetry are 0.07 mm and 10 mm for 

the skin and body, respectively. These are noted as Hp(0.07) and Hp(10), respectively. 

photon

A unit or particle of electromagnetic radiation consisting of X- or gamma rays. 

photon - X-ray

Electromagnetic radiation of energies between 10 keV and 100 keV whose source can be an 

X-ray machine or radioisotope.

pit

Nuclear weapon core, made of fissionable material.

quality factor, Q

A modifying factor used to derive dose equivalent from absorbed dose.

radiation

Alpha, beta, neutron, and photon radiation. 

radioactivity

The spontaneous emission of radiation, generally alpha or beta particles, gamma rays, and 

neutrons from unstable nuclei.

radionuclide

A radioactive isotope of an element, distinguished by atomic number, atomic weight, and 

energy state.

rem

A unit of dose equivalent equal to the product of the number of rad absorbed and the quality 

factor. The word derives from roentgen equivalent in man.

roentgen (R or r)

A unit of exposure to gamma (or X-ray) radiation. It is defined precisely as the quantity of 

gamma (or X-) rays that will produce a total charge of 2.58 × 10-4 coulomb in 1 kg of dry air. 

An exposure of 1 R is approximately equivalent to an absorbed dose of 1 rad in soft tissue for 

higher (~>100-keV) energy photons.

shallow absorbed dose (Ds)

The absorbed dose at a depth of 0.007 cm in a material of specified geometry and 

composition.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 70 of 105

shallow dose equivalent (Hs)

Dose equivalent at a depth of 0.007 cm in tissue.

shielding

Any material or obstruction that absorbs (or attenuates) radiation and thus tends to protect 

personnel or materials from radiation.

silver shield(s)

The 1-mm-thick shields covering the film packet in early personnel film dosimeters. 

skin dose

Absorbed dose at a tissue depth of 7 mg/cm2

.

thermoluminescence

Property of a material that causes it to emit light as a result of being excited by heat.

thermoluminescent dosimeter (TLD)

A holder containing solid chips of material that when heated will release stored energy as light. 

The measurement of this light provides a measurement of absorbed dose. 

whole-body dose

Commonly defined as the absorbed dose at a tissue depth of 1.0 cm (1,000 mg/cm2

); 

however, also used to refer to the recorded dose.

X-ray

Ionizing electromagnetic radiation of external nuclear origin.
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 1 of 16

LIST OF FIGURES

FIGURE TITLE

A-1 Occupational Dose Report reviewed 6-4-02, page 1 ..............................................................

PAGE

72

A-2 Occupational Dose Report reviewed 6-4-02, page 2 .............................................................. 73

A-3 Dosimetry History by Individual query report dated 3-10-03, page 1 ...................................... 74

A-4 Dosimetry History by Individual query report dated 3-10-03, page 2 ...................................... 75

A-5 Dosimetry History by Individual query report dated 3-10-03, page 3 ...................................... 76

A-6 Health Physics Yearly External Exposure Activity Run, 1953 to 1958 .................................... 77

A-7 Health Physics Yearly External Exposure Run, 1959 to 1963 ................................................ 78

A-8 Health Physics Yearly External Exposure Activity Run, 1964................................................. 79

A-9 External Dosimetry Detail, 1981 to 1983 ................................................................................ 80

A-10 Health Physics External Radiation Exposure Report, 1967 .................................................... 81

A-11 Radiation Dosimetry Individual Lifetime Report dated 6-4-02................................................. 82

A-12 Radiation Health Records System – TLD Data....................................................................... 83

A-13 Radiation Dosimetry Detail Report, Termination Report dated 9-17-96, page 1 ..................... 84

A-14 Radiation Dosimetry Detail Report, Termination Report dated 9-17-96, page 2 ..................... 85

A-15 Occupational Radiation Exposure Information, 1988.............................................................. 86
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 2 of 16

Figure A-1. Occupational Dose Report reviewed 6-4-02, page 1.
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 3 of 16

Figure A-2. Occupational Dose Report reviewed 6-4-02, page 2.
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 4 of 16

Figure A-3. Dosimetry History by Individual query report dated 3-10-03, page 1.
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 5 of 16

Figure A-4. Dosimetry History by Individual query report dated 3-10-03, page 2.
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 6 of 16

Figure A-5. Dosimetry History by Individual query report dated 3-10-03, page 3.
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 7 of 16

Figure A-6. Health Physics Yearly External Exposure Activity Run, 1953 to 1958.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 78 of 105

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 8 of 16

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

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 9 of 16

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

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 10 of 16

Figure A-9. External Dosimetry Detail, 1981 to 1983.
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 11 of 16

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

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 12 of 16

Figure A-11. Radiation Dosimetry Individual Lifetime Report dated 6-4-02.
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 13 of 16

Figure A-12. Radiation Health Records System – TLD Data.
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ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 14 of 16

Figure A-13. Radiation Dosimetry Detail Report, Termination Report dated 9-17-96, page 1.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 85 of 105

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 15 of 16

Figure A-14. Radiation Dosimetry Detail Report, Termination Report dated 9-17-96, page 2.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 86 of 105

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 16 of 16

Figure A-15. Occupational Radiation Exposure Information, 1988.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 87 of 105

ATTACHMENT B

MAJOR JOB CATEGORIES

Page 1 of 2

Chemical Operators 

Primary job duties included 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 

generally worked side-by-side with production operators.
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ATTACHMENT B

MAJOR JOB CATEGORIES

Page 2 of 2

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.
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 89 of 105

TABLE OF CONTENTS

SECTION TITLE

C.1 General Approach..................................................................................................................

PAGE

90

C.2 Applications and Limitations................................................................................................... 90

C.3 Coworker Data Development ................................................................................................. 90

C.4 Adjustment for Missed Dose .................................................................................................. 91

C.4.1 Plutonium Workers (Before 1971)...............................................................................91

C.4.2 Plutonium (After 1970) and Uranium (all years) Workers............................................ 91

C.5 Special Considerations .......................................................................................................... 92

C.6 Coworker Annual Dose Summaries ....................................................................................... 93

C.7 Plutonium Workers................................................................................................................. 93

C.7.1 1952 to 1970 ..............................................................................................................93

C.7.2 1971 to 2005 ..............................................................................................................94

C.8 Uranium Workers ...................................................................................................................95

C.9 Coworker Dose Tables........................................................................................................... 97

LIST OF TABLES

TABLE TITLE

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

PAGE

92

C-2 Lognormal neutron-to-photon ratio values, 1970 to 1976 ....................................................... 93

C-3 Annual external coworker doses for plutonium workers, 1952 to 1970 ................................... 97

C-4 Annual external coworker doses for plutonium workers, 1971 to 2005 ................................... 99

C-5 Annual external coworker doses for uranium workers, 1952 to 2005.................................... 100

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

with ORAUT-OTIB-0052....................................................................................................... 101

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

with ORAUT-OTIB-0052....................................................................................................... 104

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

with ORAUT-OTIB-0052....................................................................................................... 104

ATTACHMENT C

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT

Page 1 of 17
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 90 of 105

C.1 GENERAL APPROACH

As described in ORAUT-OTIB-0020 (ORAUT 2008), 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 had the worker been monitored.

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 TIBs for 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 prior to 1967 are not covered in this attachment. In conjunction with SEC 

Evaluation Report SEC-00032, neutron coworker doses cannot be created that would ensure that a 

bounding exposure is assigned. Therefore, neutron coworker doses should not be assigned prior to 

1967.

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. 

ATTACHMENT C

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT

Page 2 of 17
Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 91 of 105

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, External Dose Reconstruction Implementation Guideline (NIOSH 2007a), 

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. In addition, reported dose values less than one-half 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 the U.S. Department of Labor. Therefore, in the Interactive 

RadioEpidemiological Program (IREP), Parameter 1 input 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

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, then 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.

Plutonium Workers (Before 1971)

C.4.2

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.

Plutonium (After 1970) and Uranium (all years) Workers

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Table C-1. Missed external doses (rem).

Monitoring 

period

Penetrating 

LOD

Non-penetrating 

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 yr 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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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 NA NA NA

1971 1.61 3.45 12.4 NA NA NA

1972 1.32 2.15 4.64 0.8 2.63 3.94

1973 1.32b 2.15 4.64 NA NA NA

1974 0.68 3.01 4.16 NA NA NA

1975 0.67 3.31 4.82 NA NA NA

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. 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

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. 

1952 to 1970

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.

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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 

the use of the methodologies in Section 7.0 of ORAUT-OTIB-0020 (ORAUT 2008). 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 2007c). 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

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-Office 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.

1971 to 2005

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).

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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.

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 2008). 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, Parameters to Consider 

When Processing Claims for Construction Trade Workers (ORAUT 2007c). 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 

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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.

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 2008). 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, Parameters to Consider 

When Processing Claims for Construction Trade Workers (ORAUT 2007c). 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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C.9 COWORKER DOSE TABLES 

Table C-3. Annual external coworker doses for plutonium workers,

1952 to 1970 (rem).

Building Year

Photon 

50%-tile 

(rem/yr)

Photon 

95%-tile 

(rem/yr)

Neutron 

50%-tile 

(rem/yr)

Neutron 

95%-tile 

(rem/yr)

71 1952 N/A

N/A

71 1953 0.216 1.752

71 1954 0.520a 3.340

71 1955 0.469 6.518

71 1956 0.456 4.890

71 1957 0.562 5.371

71 1958 0.556 6.309

71 1959 1.356 6.779

71 1960 0.574 3.546

71 1961 1.330 4.849

71 1962 1.069 4.105

71 1963 0.574 3.337

71 1964 0.261 2.320

71 1965 0.523 4.000

71 1966 1.192 6.037

71 1967 0.907 5.614 0.808 3.941

71 1968 0.678 3.468 0.991 6.935

71 1969 0.590 4.278 1.590 4.824

71 1970 0.120a 4.278b 1.675 5.110

76 1952

N/A

N/A

76 1953

76 1954

76 1955

76 1956

76 1957 0.520a 2.699

76 1958 0.730 6.361

76 1959 1.616 6.153

76 1960 1.981 5.840

76 1961 2.451 5.319

76 1962 1.721 4.484

76 1963 1.616 4.563

76 1964 1.741 5.079

76 1965 0.913 5.342

76 1966 1.898 7.324

76 1967 0.706 6.486 0.971 2.774

76 1968 0.283 2.759 0.954 3.441

76 1969 0.065 1.570 1.217 2.292

76 1970 0.120a 1.102 1.703 2.738

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Building Year

Photon 

50%-tile 

(rem/yr)

Photon 

95%-tile 

(rem/yr)

Neutron 

50%-tile 

(rem/yr)

Neutron 

95%-tile 

(rem/yr)

77 1952

N/A

N/A

77 1953

77 1954

77 1955

77 1956

77 1957 0.480a 0.480a

77 1958 0.052 2.918

77 1959 0.261 2.868

77 1960 0.130 3.598

77 1961 1.121 5.214

77 1962 0.479 2.764

77 1963 0.815 4.343

77 1964 0.965 4.958

77 1965 0.754 3.682

77 1966 1.278 5.519

77 1967 1.454 5.299 1.048 1.966

77 1968 0.586 2.772 0.969 1.811

77 1969 0.088 1.991 0.717 1.325

77 1970 0.120a 2.392 2.060 2.873

91 1952 0.782 1.646

N/A

91 1953 0.332 2.257

91 1954 0.520a 0.782

91 1955 0.520a 0.913

91 1956 0.480a 0.521

91 1957 0.520a 0.521

91 1958 0.480a 0.521

91 1959 0.240a 0.521

91 1960 0.240a 0.261

91 1961 0.240a 0.521

91 1962 0.240a 0.519

91 1963 0.240a 0.587

91 1964 0.240a 0.417

91 1965 0.240a 0.706

91 1966 0.118 1.662

91 1967 0.240a 1.276 1.189b 1.669b

91 1968 0.240a 0.665 0.669 1.073

91 1969 0.120a 0.483 1.108 2.237

91 1970 0.209 1.322 1.108b 2.237b

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Building Year

Photon 

50%-tile 

(rem/yr)

Photon 

95%-tile 

(rem/yr)

Neutron 

50%-tile 

(rem/yr)

Neutron 

95%-tile 

(rem/yr)

All Buildings 1952 0.782 1.646

N/A

All Buildings 1953 0.261 1.947

All Buildings 1954 0.520a 1.825

All Buildings 1955 0.240a 4.596

All Buildings 1956 0.130 3.859

All Buildings 1957 0.480a 3.702

All Buildings 1958 0.140 4.355

All Buildings 1959 0.782 5.788

All Buildings 1960 0.433 4.745

All Buildings 1961 1.051 5.058

All Buildings 1962 0.730 3.964

All Buildings 1963 0.548 3.807

All Buildings 1964 0.339 3.285

All Buildings 1965 0.393 3.852

All Buildings 1966 0.900 5.958

All Buildings 1967 0.482 5.137 0.873 3.134

All Buildings 1968 0.281 2.869 0.965 5.044

All Buildings 1969 0.135 2.997 1.348 4.356

All Buildings 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

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1992 0.188 0.105 0.079 0.044 5,831

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

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

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Year

Penetrating

95th percentile

Penetrating

50th percentile

Nonpenetrating

95th percentile

Nonpenetrating

50th percentile

Number of 

monitored workers

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)

(ORAUT 2007c).

Building Year

Photon 

50%-tile 

(rem/yr)

Photon 

95%-tile 

(rem/yr)

Neutron 

50%-tile 

(rem/yr)

Neutron 

95%-tile 

(rem/yr)

71 1952 N/A

N/A

71 1953 0.303 2.453

71 1954 0.728a 4.676

71 1955 0.657 9.125

71 1956 0.639 6.846

71 1957 0.786 7.519

71 1958 0.779 8.833

71 1959 1.898 9.490

71 1960 0.803 4.964

71 1961 1.862 6.789

71 1962 1.497 5.747

71 1963 0.803 4.672

71 1964 0.365 3.249

71 1965 0.732 5.600

71 1966 1.669 8.451

71 1967 1.269 7.860 1.132 5.518

71 1968 0.949 4.855 1.387 9.709

71 1969 0.825 5.989 2.227 6.754

71 1970 0.168a 5.989b 2.345 7.154

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Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 102 of 105

Building Year

Photon 

50%-tile 

(rem/yr)

Photon 

95%-tile 

(rem/yr)

Neutron 

50%-tile 

(rem/yr)

Neutron 

95%-tile 

(rem/yr)

76 1952

N/A

N/A

76 1953

76 1954

76 1955

76 1956

76 1957 0.728a 3.778

76 1958 1.022 8.906

76 1959 2.263 8.614

76 1960 2.774 8.176

76 1961 3.431 7.446

76 1962 2.409 6.278

76 1963 2.263 6.388

76 1964 2.437 7.110

76 1965 1.278 7.479

76 1966 2.657 10.254

76 1967 0.989 9.081 1.360 3.884

76 1968 0.396 3.863 1.335 4.818

76 1969 0.091 2.197 1.703 3.209

76 1970 0.168a 1.543 2.385 3.833

77 1952

N/A

N/A

77 1953

77 1954

77 1955

77 1956

77 1957 0.672a 0.672a

77 1958 0.073 4.086

77 1959 0.365 4.015

77 1960 0.183 5.037

77 1961 1.570 7.300

77 1962 0.670 3.869

77 1963 1.141 6.080

77 1964 1.351 6.942

77 1965 1.055 5.155

77 1966 1.789 7.726

77 1967 2.036 7.419 1.467 2.753

77 1968 0.820 3.881 1.357 2.535

77 1969 0.123 2.787 1.004 1.855

77 1970 0.168a 3.349 2.885 4.022

ATTACHMENT C

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT

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Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 103 of 105

Building Year

Photon 

50%-tile 

(rem/yr)

Photon 

95%-tile 

(rem/yr)

Neutron 

50%-tile 

(rem/yr)

Neutron 

95%-tile 

(rem/yr)

91 1952 1.095 2.304

N/A

91 1953 0.465 3.160

91 1954 0.728a 1.095

91 1955 0.728a 1.278

91 1956 0.672a 0.730

91 1957 0.728a 0.730

91 1958 0.672a 0.730

91 1959 0.336a 0.730

91 1960 0.336a 0.365

91 1961 0.336a 0.730

91 1962 0.336a 0.727

91 1963 0.336a 0.821

91 1964 0.336a 0.584

91 1965 0.336a 0.989

91 1966 0.165 2.327

91 1967 0.336a 1.786 1.665b 2.337b

91 1968 0.336a 0.931 0.936 1.503

91 1969 0.168a 0.676 1.551 3.131

91 1970 0.292 1.850 1.551b 3.132b

All Buildings 1952 1.095 2.304

N/A

All Buildings 1953 0.365 2.725

All Buildings 1954 0.728a 2.555

All Buildings 1955 0.336a 6.435

All Buildings 1956 0.183 5.402

All Buildings 1957 0.672a 5.183

All Buildings 1958 0.197 6.097

All Buildings 1959 1.095 8.103

All Buildings 1960 0.606 6.643

All Buildings 1961 1.472 7.081

All Buildings 1962 1.022 5.549

All Buildings 1963 0.767 5.330

All Buildings 1964 0.475 4.599

All Buildings 1965 0.551 5.393

All Buildings 1966 1.260 8.341

All Buildings 1967 0.675 7.192 1.223 4.388

All Buildings 1968 0.393 4.016 1.352 7.061

All Buildings 1969 0.189 4.195 1.887 6.098

All Buildings 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.

ATTACHMENT C

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT

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Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 104 of 105

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

accordance with ORAUT-OTIB-0052 (rem)(ORAUT 2007c).

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

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 2007c).

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

ATTACHMENT C

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT

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Document No. ORAUT-TKBS-0011-6 Revision No. 02 Effective Date: 08/14/2007 Page 105 of 105

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

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

ATTACHMENT C

EXTERNAL COWORKER DOSIMETRY DATA FOR ROCKY FLATS PLANT

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