ORAU TEAM

Dose Reconstruction

Project for NIOSH

 

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

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Document Title:

Rocky Flats Plant – Occupational External Dose

Document Number: ORAUT-TKBS-0011-6 

Revision: 01 

Effective Date: 02/08/2007

Type of Document: TBD

Supersedes: Revision 00 

Subject Expert(s): James M. Langsted

Site Expert(s): N/A

Approval: Signature on File Approval Date: 01/29/2007

Robert Meyer, Document Owner

Approval: Signature on File Approval Date: 01/19/2007

John M. Byrne, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 01/18/2007

Edward F. Maher, Task 5 Manager

Concurrence: Signature on File Concurrence Date: 01/25/2007

James P. Griffin, Deputy Project Director

Approval: Brant A. Ulsh Signature on File for Approval Date: 02/08/2007

James W. Neton, Associate Director for Science
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 2 of 90

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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 3 of 90

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............................................................................ 13

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

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

6.3.2.1 Rounding ..................................................................................................... 16

6.3.2.2 Deep Dose Not Equal to Gamma Plus Neutron Doses................................. 16

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

6.3.3.1 RHRST_ED_TLD_HISTORY ....................................................................... 17

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

6.3.4 Interpretation of Dosimetry Data.................................................................................20

6.3.5 Additional Data Available............................................................................................20

6.3.5.1 Rocky Flats Work History File ...................................................................... 20

6.3.5.2 Neutron Dose Reconstruction Project File.................................................... 24

6.3.5.3 Job Exposure Matrix .................................................................................... 24

6.4 Historical Administrative Practices ......................................................................................... 24

6.4.1 Badged Population .....................................................................................................24

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...............................................................................................................33

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

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

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
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 4 of 90

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

6.7 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 ............................. 46

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

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

6.7.3.2 Number of Zero Readings ............................................................................ 48

6.7.3.3 Unmonitored Energy Range......................................................................... 48

6.7.3.4 Neutron Dose Reconstruction Project .......................................................... 49

6.7.3.5 Default Neutron-to-Gamma Ratio................................................................. 50

6.7.4 Geometry....................................................................................................................51

6.7.4.1 Angular Dependence ................................................................................... 51

6.7.4.2 Exposure Geometry ..................................................................................... 51

6.7.5 Uncertainty .................................................................................................................51

6.7.5.1 Film..............................................................................................................51

6.7.5.2 Thermoluminescent Dosimeter..................................................................... 52

6.8 Electron Dose ........................................................................................................................52

6.8.1 Energy Groups ...........................................................................................................53

6.8.1.1 Exposure Spectra......................................................................................... 53

6.8.1.2 Reported Dose to Energy Groups ................................................................ 54

6.8.2 Calibration Factor .......................................................................................................54

6.8.2.1 Reported-Dose-to-Organ-Dose Conversion Factor Units ............................. 54

6.8.3 Missed Dose...............................................................................................................54

6.8.3.1 Limit of Detection ......................................................................................... 54

6.8.3.2 Number of Zero Readings ............................................................................ 55

6.8.3.3 Unmonitored Energy Range......................................................................... 55

6.8.4 Geometry....................................................................................................................56

6.8.4.1 Angular Dependence ................................................................................... 56

6.8.4.2 Exposure Geometry ..................................................................................... 56

6.8.5 Uncertainty .................................................................................................................56

6.8.5.1 Film..............................................................................................................56

6.8.5.2 Thermoluminescent Dosimeter..................................................................... 57

6.8.6 Skin Contamination ....................................................................................................58

6.9 Unmonitored Individuals......................................................................................................... 59

6.9.1 In Production Areas ....................................................................................................59

6.9.2 Outside Production Areas...........................................................................................60
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 5 of 90

6.10 Extremity Dosimetry...............................................................................................................60

6.11 Attributions and Annotations .................................................................................................. 61

References ........................................................................................................................................65

Glossary ............................................................................................................................................69

ATTACHMENT A, EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS........................... 73

ATTACHMENT B, MAJOR JOB CATEGORIES ................................................................................ 89
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 6 of 90

LIST OF TABLES

TABLE TITLE

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

PAGE

14

6-2 Interpretation of reported data................................................................................................ 21

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........................................................... 44

6-16 Neutron dose measurements divided into energy groups....................................................... 45

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 ........................................................................................... 47

6-20 Neutron LODs........................................................................................................................48

6-21 Potential missed neutron dose for early film dosimeters......................................................... 48

6-22 Correction ratios for identified neutron film reading deficiencies............................................. 50

6-23 NDRP-developed neutron-to-gamma ratios ........................................................................... 50

6-24 ORAU Team-developed neutron-to-gamma ratios ................................................................. 51

6-25 Uncertainty for TLD neutron dose measurements .................................................................. 52

6-26 Beta LODs .............................................................................................................................55

6-27 Uncertainty for beta film readings........................................................................................... 57

6-28 Uncertainty for loose-chip TLD beta dose .............................................................................. 57

6-29 Uncertainty for DOELAP-accredited TLD beta dose............................................................... 58

6-30 DU mixtures ...........................................................................................................................59

6-31 One-year-old DU....................................................................................................................60
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 7 of 90

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...................................................................................................... 13

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 ........................... 53

6-8 Shallow dose rate from natural uranium slab.......................................................................... 54

A-1 Occupational Dose Report reviewed 6-4-02, page 1 .............................................................. 74

A-2 Occupational Dose Report reviewed 6-4-02, page 2 .............................................................. 75

A-3 Dosimetry History by Individual query report dated 3-10-03, page 1 ...................................... 76

A-4 Dosimetry History by Individual query report dated 3-10-03, page 2 ...................................... 77

A-5 Dosimetry History by Individual query report dated 3-10-03, page 3 ...................................... 78

A-6 Health Physics Yearly External Exposure Activity Run, 1953 to 1958 .................................... 79

A-7 Health Physics Yearly External Exposure Run, 1959 to 1963 ................................................ 80

A-8 Health Physics Yearly External Exposure Activity Run, 1964................................................. 81

A-9 External Dosimetry Detail, 1981 to 1983 ................................................................................ 82

A-10 Health Physics External Radiation Exposure Report, 1967 .................................................... 83

A-11 Radiation Dosimetry Individual Lifetime Report dated 6-4-02................................................. 84

A-12 Radiation Health Records System – TLD Data....................................................................... 85

A-13 Radiation Dosimetry Termination Report dated 9-17-96, page 1............................................ 86

A-14 Radiation Dosimetry Termination Report dated 9-17-96, page 2............................................ 87

A-15 Occupational Radiation Exposure Information, 1988.............................................................. 88
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 8 of 90

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
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 9 of 90

MeV megaelectron-volt, 1 million electron-volts

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

NTA nuclear track emulsion, type A

ORNL Oak Ridge National Laboratory

ORAU Oak Ridge Associated Universities 

OSL Optically Stimulated Luminescence

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

SOE stationary operating engineer

SRDB Site Research Database

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. 01 Effective Date: 02/08/2007 Page 10 of 90

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

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

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

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

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

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

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

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

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

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

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

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

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

derived:

] guidelines established under 

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

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

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

• Radiation from naturally occurring radon present in conventional structures

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

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 11 of 90

6.1.1

This 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 2006). 

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 (2006), this document identifies how to 

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

protection factors. 

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). 

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 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 12 of 90

Figure 6-1. ORNL-style film badge 

(including brass filter).

Figure 6-2. RFP multielement film badge.

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). 

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

Chemical Company (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. 

Figure 6-3. RFP interim TLD/film 

badge.

Figure 6-4. RFP Harshaw badge.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 13 of 90

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 neutron-sensitive elements 

and one neutron-insensitive element 

under cadmium or tin filters. This 

badge includes a PNAD (Baker 

2002). Figure 6-5. RFP Panasonic dosimeter.

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 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. 

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 provides 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). 
Document No. ORAUT-TKBS-0011-6 Revision No. 01

Effective Date: 02/08/2007 Page 14 of 90

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. 01

Effective Date: 02/08/2007 Page 15 of 90

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. 01 Effective Date: 02/08/2007 Page 16 of 90

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 

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 on the report cards of 1 mrem per 

measurement. 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 Possible Algorithm Issue

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

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

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

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

electronic record was zero.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 17 of 90

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 Possible Manual Correction

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

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

review of a number of these records 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

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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 18 of 90

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 an identical structure to RHRST_ED_TLD_HISTORY, contains post-1991 data. 

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

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

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

activity date.

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

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

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

employment period. This is particularly true for subcontractors.

Dose History 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]:
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 19 of 90

• 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

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 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 20 of 90

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 Detail 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)

• 1977–2005, dosimeter exchange history

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

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

claimant's external dosimetry record.

6.3.4

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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 21 of 90

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. 01 Effective Date: 02/08/2007 Page 22 of 90

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. 01 Effective Date: 02/08/2007 Page 23 of 90

 

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. 01 Effective Date: 02/08/2007 Page 24 of 90

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), and the use of these data in dose reconstructions is detailed in ORAUT (2005a).

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.

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. 

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 

 

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. 01 Effective Date: 02/08/2007 Page 25 of 90

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. 

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

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

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

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

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

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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 26 of 90

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 ORAU 2006d). 

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 

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

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 the author of this report, 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 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 27 of 90

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 

Minimum Reported Dose
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 28 of 90

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. 01 Effective Date: 02/08/2007 Page 29 of 90

was reported as zero (RFETS 2001). This policy is consistent with the LODs reported elsewhere in 

this TBD.

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. 01 Effective Date: 02/08/2007 Page 30 of 90

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 must 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 2006).

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 (2006), 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 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 31 of 90

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. 01 Effective Date: 02/08/2007 Page 32 of 90

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

reconstruction.

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).

Missing Entry

6.5.4

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

reconstructions is favorable to claimants. An alternative approach is presented below.

Exposure Geometry

Because little information is available on the exposure geometry for an individual, estimates have 

been made by the author using professional judgment (NIOSH 2006, 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%
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 33 of 90

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%

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 

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. 

Lead Aprons

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
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 34 of 90

Although this field study was performed using Panasonic dosimeters, the use of 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 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. 

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

dosimeters.

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 indicates 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 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 35 of 90

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

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 highbackground 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 - February 2000) standard 

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

Technical Basis Document for Rocky Flats Plant Occupational Environmental Dose (ORAU 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). This may need to be addressed for best 

estimate Dose Reconstructions.

6.6 PHOTON DOSE

6.6.1

The NIOSH IREP software for calculating the POC (NIOSH 2006) contains three photon energy 

bands:

Energy Groups
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 36 of 90

• Below 30 keV

• 30 to 250 keV

• Above 250 keV 

Separation of the dose from each energy band is required.

6.6.1.1 Default Exposure Spectra

Very little spectroscopy data that indicate the gamma spectrum in RFP work areas have been found. 

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 the algorithms in 

ORAUT (2005b). 

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 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 37 of 90

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.

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]. 

Use of this data in dose reconstructions is detailed in ORAUT (2005b).

Three-Element Dosimeter 

As indicated in Section 6.2, the three-element film dosimeter used at RFP from 1960 to 1969 had an 

open window (dose indicated under OW in the reports), a 1-mm thick cadmium (CD) filter, and a 

brass (BR) filter that provided half the filtration of the CD filter. The brass filter was added to measure 

more accurately the 60-keV photons. 

Some dosimetry records (illustrated in Figure A-7) indicate the dose determined by 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 (from the plutonium areas), the skin dose was calculated as a sum of 

each of the three windows (OW + CD + BR). The penetrating dose was calculated by adding CD, BR, 

and 35% of the OW readings. 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. Different algorithms were used for 

uranium area beta exposures (Table 6-4).

To reconstruct the plutonium area low- and intermediate-energy photon dose between 1960 and 1969 

properly, the following reverse algorithm should be applied:

(6-1)

(6-2)

This method effectively recreates the measured dose from under the BR and CD filters, which 

represent the 30- to 250-keV dose. An alternative approach to determining the intermediate-energy 

(30- to 250-keV) photon dose is simply to sum the BR and CD doses, if available in the energy 

employee's dose record submitted by DOE. The low-energy (<30-keV) photon dose can be 

determined by simply using the OW dose reading.

Two-Element Dosimeter 

The RFP two-element dosimeter used before 1960 effectively could not measure the 60-keV photons; 

therefore, the penetrating dose for the plutonium areas was calculated using Pen = 50% OW + CD. 

Based on the BR to CD ratio observed in plutonium worker dosimetry data that was reviewed, this 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 38 of 90

method is surprisingly accurate in determining the dose contribution from 30- to 250-keV photons 

(96% in comparison to the three-element dosimeter). The disadvantage is that using only the 

penetrating dose will underestimate the total photon dose (low and intermediate energy). As a result, 

the following method should be used to estimate the dose from low-and intermediate-energy photons: 

(6-3)

(6-4)

At first glance, it could appear that these calculations overestimate the low-energy dose from the 

plutonium areas (i.e., part of the skin dose is being counted twice, so the sum of the low-and 

intermediate-energy doses exceeds the original skin dose). This effect results because the 

intermediate-energy photon dose was underestimated due to the absence of the 60-keV photon dose 

contribution. The original reported skin dose is actually underestimated. The addition of 50% of the 

OW dose to the penetrating dose appears to correct the Pen for this underestimation (i.e., the relative 

ratio of the 60-keV photon dose to the OW dose contribution is about 0.5). Review of the DuPont 

energy response curves for Type 502 and 508 films indicates that the densitometer difference 

between exposure to 17-keV photons and 60-keV photons is significant. It is believed that the lowenergy photon film response dominated the OW densitometer measurements, and the 60-keV 

photons were therefore not effectively measured.

Comparing these two methods on two plutonium worker data sets, the calculated low- to intermediateenergy dose ratio is approximately the same at approximately 0.8. The skin-to-penetrating dose ratio 

is also similar at around 0.7, as well as the OW/CD dose ratio at about 1.45. As a result, it is believed 

that the methods outlined above accurately reflect the plutonium area low- (<30-keV) and 

intermediate-energy (30- to 250-keV) dose contributions.

After 1968, RFP used TLDs to measure photon and beta dose. The TLD materials used were much 

more tissue-equivalent and the response much less energy-dependent. Dosimeters were calibrated 

to more appropriate radiation energies, and filter design had advanced. It is believed that these 

dosimeters performed substantially better than film.

It is reasonable to assume that for plutonium workers, the skin dose includes the shallow dose from 

photons. For uranium workers, the skin dose can be assumed to include electrons.

The measured results trend of the dosimeters worn at Rocky Flats Plant is shown in Section 6.4.1 

(Figure 6-6a).

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 2005c, 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 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 39 of 90

dosimetry system, which was performed using DOE Laboratory Accreditation Program (DOELAP) 

sources at Pacific Northwest Laboratory (PNL). DOELAP sources have been used since that time. 

The personal dose equivalent [Hp(10)] is the appropriate unit to use for this period. Table 6-11 

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 2006. 

Plutonium-specific photon dose conversion factors are provided in Table 4.1a of NIOSH 2006 and 

should be applied for plutonium exposures at RFP.

6.6.3

Section 2.1.2 of NIOSH (2006) recommends the use of the limit of detection (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 less sensitive DuPont 1290 film. The 1290 film was not processed unless the 

580 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 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 by the author 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 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 40 of 90

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.

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 NIOSH (2006). 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 the dose reconstruction guidance discuss this estimate. In this case, the estimate 

is assumed to have a lognormal distribution (NIOSH 2006, 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 2005d) 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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 41 of 90

The initial implementation of the Panasonic TLD system was based on a range of DOELAP exposure 

categories. The response of the dosimeter was evaluated in relation to these exposures, and the 

algorithm was derived from these exposures. Therefore, the initial implementation of the Panasonic 

TLD system and the later DOELAP-accredited operation of that system are unlikely to have missed 

photon dose in an energy range to which workers could be exposed. No missed-dose correction is 

appropriate for this dosimetry system.

6.6.4

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

Exposure geometry is common to all types of radiation exposure, as addressed in Section 6.5.4.

6.6.5

The External Dose Reconstruction Guideline (NIOSH 2006) 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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 42 of 90

The method in NIOSH (2006) 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 

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

6.6.5.2 Thermoluminescent Dosimeter

TLDs provided improved photon dosimetry. This section estimates the uncertainty associated with 

this type of dosimeter for the early years of use and then discusses the measured uncertainties after 

1983 when DOELAP performance testing began.

6.6.5.2.1 Loose-Chip Thermoluminescent Dosimeters

Harshaw TLD chips were used to measure photon dose at RFP from 1969 to 1982. These chips 

were carried in a dosimeter holder but were removed to be read (thus the term loose). A calculation 

was performed to estimate the uncertainty associated with reading the photon dose from these 

dosimeters.

Little information has been found that describes the variability of response when these chips were in 

service. A chip-sorting procedure was used to remove chips from service that had responded outside 

set criteria (Link and Pennock 1983). The procedure was to expose the chips to a 1,000-mrem dose 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 43 of 90

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) ÷ 

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 (2006), 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. 

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 Panasonic Thermoluminescent Dosimeter

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 (2006). 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].
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 44 of 90

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.

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 (2006). 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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 45 of 90

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

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].

Table 6-17. Default neutron energy distribution. 

Neutron energy 

intervals

Fraction of dose 

(NCRP 38)

Dose multiplier 

(ICRP 60) Dose multipliera

<10 keV 0.035 2.13 0.0755

10–100 keV 0.017 1.86 0.0309

0.1–2 MeV 0.687 1.91 1.31

2.0–20 MeV 0.261 1.32 0.345

>20 MeV 0 None None

a. Multiply the reported dose by these factors to determine the ICRP 60 neutron dose for 

each neutron energy interval. 

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

NIOSH (2006) 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 

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

measurements and the ICRP (1991) radiation weighting factors is 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 2006.
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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 2005c, 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 National Bureau of 

Standards Handbook 63 (NBS 1957). A set of polyethylene moderators was constructed. The 

spectra from these moderated sources compared well with work area spectra measured with a 

precision long counter and a series of paraffin moderators fitted over the counter (Mann and Boss 

1963). Ambient dose equivalent [H*(10)] is appropriate for this dosimeter.

Harshaw TLDs at RFP were initially calibrated using a 210-g PuF4 source built at RFP and calibrated 

at the LANL standard pile, which was established as a neutron flux standard (Mann and Boss 1963). 

A set of polyethylene moderators was constructed to provide various degrees of moderation. The 

bare PuF4 source dose rate was calculated using neutron spectra from an unknown reference 

document and quality factors published in DOE Orders (Falk 1975). The dose rates for the 

moderated spectra were measured with currently available neutron dose rate instrumentation. The 

PuF4 source was placed in storage in about 1975 and replaced with a commercially manufactured 

and calibrated 252Cf source. The calculation of the dose rate used a published spectrum and dose 

rate (Barker 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 (2005c). 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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 47 of 90

Beginning in July 1958, RFP processed NTA film at the site. The NDRP (ORAUT 2005a) 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. 

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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 48 of 90

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 Unmonitored Energy Range

NTA film is a poor detector of neutron energies below 500 to 800 keV (Griffith et al. 1979; Wilson et 

al. 1990). Before 1963, RFP appears to have calibrated neutron film with a variety of unmoderated 

neutron sources. RFP recognized that dosimetry results were not consistent with instrument 

measurements and that low-energy neutrons were not measured by the film. A project was initiated 

in 1962 to improve neutron film dosimeter calibration (Mann and Boss 1963).

Before 1963, neutron dose from neutrons below approximately 800 keV probably was not detected. 

To determine how much dose was potentially missed, the neutron measurements performed in RFP 

work areas (Brackenbush et al. 1989) were corrected for ICRP (1991) radiation weighting factors and 

the fraction of the dose from neutrons of less than 800 keV was determined. Table 6-21 lists these 

values.

Table 6-21. Potential missed neutron dose for early film 

dosimeters. 

Location

ICRP 60

rem/hr

Below

800 keV

Building 771 fluorinator line 1.13E-03 52%

Building 776 molten salt glovebox 3.25E-03 60%

Building 776 molten salt storage vault 1.67E-02 29%

Building 776 drum storage 4.46E-02 57%

Building 707 high dose pit 1.18E-03 22%

Building 707 low dose pit 5.26E-04 29%

Building 707 oxide can 2.53E-03 47%

Building 707 plutonium ingot 3.70E-03 16%

There appears to be no distinct pattern in these data. It is appropriate to take an approach favorable 

to claimants and select the largest value of 60%. Therefore, the total neutron dose from RFP 

measurements before 1964 (1951 to 1963) should be multiplied by 2.5 before applying the factors 

from Table 6-17.

In 1962, RFP began a project to refine neutron dosimeter calibration to match the neutron spectra in 

the production areas more accurately. Mann and Boss (1963) documented an effort to develop a 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 49 of 90

calibrated PuF4 source with various moderators. The spectra from the moderator configurations of 

this source were compared with neutron spectra measurements taken in the plutonium production 

areas with a precision long counter and a series of paraffin moderators. This resulted in dosimeter 

calibrations that more accurately matched the exposure spectra. No missed dose correction is 

required for RFP neutron film dosimeters after 1963.

The RFP TLD neutron dosimeter systems (Harshaw and Panasonic) were calibrated using variously 

moderated spectra. There is no need for missed neutron dose corrections. After 1990, the 

Panasonic TLD system was DOELAP-accredited, which supports the decision to forego a missed 

neutron dose correction.

6.7.3.4 Neutron Dose Reconstruction Project

In the early 1990s, RFP addressed the issue of neutron film processing. It had been long recognized 

that, in the dosimetry laboratory, human factors associated with reading large numbers of neutron 

films under a microscope can significantly affect neutron dosimetry results. A pilot study in 1994 

reevaluated neutron doses for selected plutonium workers. This study indicated that the original 

evaluations of films might have contained significant errors, and that the resultant neutron doses 

might be significantly higher or lower than the doses actually received. The NDRP was initiated to 

provide current and former radiation workers an assessment of the neutron exposure received in the 

plutonium production facilities. The scope of this project covered 1952 to 1970.

Two methods were used to identify workers for evaluation by the NDRP. The initial method was 

identification of workers using the neutron dosimetry worksheets. These sheets identified those 

workers assigned neutron-sensitive elements (i.e., neutron films or glass plates). A portion of the 

neutron worksheets indicates issue of neutron dosimeters to personnel whose home building 

assignment was not a plutonium production building (such as Buildings 21, 22, 23, 34, 44, 81, and 

86). These individuals worked in non-neutron buildings but were issued neutron dosimeters because 

they occasionally performed work activities in plutonium production buildings. Examples of these job 

descriptions include guards, radiation monitors, technical researchers, and uranium process 

operators.

The second identification method was through use of the beta-gamma worksheets for plutonium 

production buildings. The analysis used only the beta-gamma worksheets from the plutonium 

production buildings (any building with a number starting with 7), Buildings 91 and 86, and the 

combined worksheets for Buildings 21, 22, and 23. The rosters from the beta-gamma worksheets for 

these buildings were used to identify workers to be assigned a notional neutron dose by the NDRP if 

they were not monitored for neutrons. Beta-gamma worksheets for other buildings were not used.

The NDRP reread neutron films (where available) with appropriate quality controls and reevaluated 

the neutron doses. Notional neutron doses were determined for plutonium workers with missing or 

unreadable films and for non-neutron-monitored workers in plutonium production buildings (Falk et al. 

2005).

To provide a correction favorable to claimants until NDRP data became available, it was appropriate 

to use the neutron correction ratio in Report of Epidemiologic Analyses Performed for Rocky Flats 

Production Workers Employed Between 1952-1989 (Ruttenber et al. 2003). These analyses used a 

combination of workplace instrument measurements and Harshaw and Panasonic TLD results to 

estimate correction ratios for total penetrating doses. These ratios provide an estimate for the total 

penetrating dose (gamma + neutron), which provides an initial correction for the identified bias in the 

neutron film reading. This correction should be applied to personnel who worked in the noted neutron 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 50 of 90

buildings from 1951 to 1967. When the NDRP was completed, a more accurate neutron dose 

became available for each plutonium worker who was monitored for neutrons. An updated ratio was 

generated from these data for use with unmonitored workers in this work area. Table 6-22 lists the 

initial correction ratios that were used.

Table 6-22. Correction ratios for identified neutron film 

reading deficiencies. 

Building Mean Standard deviation

771 1.99 0.92

Other neutron buildingsa 1.13 0.82

a. Buildings 123, 774, 776, 777, 779, 886, 991, and others if 

record suggests neutron monitoring

The NDRP developed neutron-to-gamma ratios for use on that project; the values are shown in 

Table 6-23, and their derivation is documented in the NDRP report (Falk et al. 2005). It is appropriate 

to use these values for those buildings and years, and other plutonium exposure situations at RFP. 

Uranium exposure is addressed below.

6.7.3.5 Default Neutron-to-Gamma Ratio

To complete the neutron-to-gamma ratio coverage for the life of the Plant, the ORAU Team 

developed geometric mean (GM) and geometric standard deviation (GSD) ratios from the available 

RFP worker files (ORAUT 2005a). These data are shown in Table 6-24, and the derivation is 

documented in the reference. As indicated in the referenced text, the overall average is appropriate 

to use from 1970 to 1976.

Table 6-23. NDRP-developed neutron-to-gamma ratios (Falk et al. 2005). 
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Table 6-24. ORAU Team-developed neutron-to-gamma ratios (ORAUT 2005a).

As discussed in Sections 6.3 and 6.3.1, RFP reported:

• Deep dose + neutron as penetrating (Pen) or deep

• Shallow dose + neutron as Skin 

In the early years, the neutron and gamma dose values were not retained and only the penetrating 

value remains. In this case, it is necessary to estimate the neutron and photon dose for dose 

reconstruction. Ruttenber et al. (2003) addressed this issue by assuming a 2:1 neutron/gamma ratio 

for Building 771, 0.5:1 for other buildings with a potential for neutron exposure (Buildings 123, 774, 

776, 777, 779, 886, 991, and other buildings if the record suggests neutron monitoring), and 0:1 for all 

other RFP facilities. This assumption is based on measurements taken at RFP and was appropriate 

to use until the NDRP was completed and provided better information for neutron-monitored 

individuals. It is appropriate to use these more refined values as they become available.

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

The worker exposure geometry for neutron dose is similar to that for photons, which is discussed in 

Section 6.5.4.

6.7.5

6.7.5.1 Film

Uncertainty

The NDRP has evaluated film uncertainty. Uncertainty for neutron film results are taken from that 

study as reported in ORAUT (2005a).
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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 (2006) 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 6-25 lists the uncertainties for these dosimetry systems.

Table 6-25. 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.

6.8 ELECTRON DOSE

Beta radiation fields are usually the dominant external radiation hazard in facilities that require contact 

work with unshielded forms of uranium. This was the case at RFP for EU and DU work. It should be 

assumed that the skin dose reported in RFP dosimetry records for uranium workers is electron 

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

natural uranium used for this experiment.

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

(DOE 2001).
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 54 of 90

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

6.8.1.2 Reported Dose to Energy Groups

NIOSH (2006) 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.

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 (2004, 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.
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Harshaw TLDs were used for beta detection starting in 1969. ORAUT (2004, Section 6.5.2) states 

that the minimum detectable dose would have been similar to that for photons. Wilson et al. (1990) 

determined that the Hanford TLD system had a 20-mR minimum detectable dose. RFP TLD 

measurements were similar. A minimum detectable dose of 20 mrem beta (shallow) is appropriate for 

RFP during this period.

The algorithm initially developed for Panasonic TLD system implementation in 1983 contains a 

constraint to ensure that the shallow dose equivalent does not fall below 0.9 times the deep dose from 

photons. Therefore, the shallow minimum detectable dose is 0.9 times that determined for deep dose 

photons (20 mrem) for this system. The minimum detectable shallow dose for this period was 

determined to be 20 × 0.9 = 18 mrem (shallow) as indicated in Table 6-26.

Table 6-26. 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 6-26.

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.
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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 (2006) 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 2006) 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 

6-27 lists these upper 95% confidence doses.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 57 of 90

Table 6-27. Uncertainty for beta 

film readings.

Dose 

(mrad)

Upper 95% confidence 

dose (mrad)

1 17

2 18

5 21

10 26

20 36

50 66

100 120

200 240

500 600

1,000 1,200

2,000 2,400

6.8.5.2 Thermoluminescent Dosimeter

TLDs provided improved beta dosimetry. Harshaw TLD chips were used to measure beta dose at 

RFP from 1969 to 1982. This section estimates the uncertainty associated with this type of dosimeter 

for the early years of use and then discusses the measured uncertainty when DOELAP performance 

testing was initiated.

6.8.5.2.1 Loose-Chip Thermoluminescent Dosimeters

Harshaw TLD chips were used to measure beta dose in parallel with photon dose. As with the photon 

TLD uncertainty, the chip-sorting procedure was used to estimate the standard error associated with 

the beta TLD measurements. Using the Simplified Dosimetry Uncertainty calculation recommended 

by NIOSH (2006), and assuming that the critical level (Lc) is the beta LOD estimated in 

Section 6.8.3.1 of this TBD, Table 6-28 lists the upper 95% confidence dose.

Table 6-28. 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

6.8.5.2.2 Panasonic TLD Dosimeter

Table 6-29 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 (2006). This method recognizes that the elements of the uncertainty are 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 58 of 90

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 

(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.

Table 6-29. 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

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

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

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 (2006). 

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. This instrument typically has a 33.3% ±1% 

(cpm/dpm) efficiency for DU. 

DU consists of 99.8% 238U by weight. Table 6-30 lists the other isotopes.
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Table 6-30. 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 6-31 lists the full set of decay isotopes. 

Dose calculation might utilize software such as VARSKIN (recommended in NIOSH 2006) 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. During this period the 

guideline was 3 rem/qtr. Therefore, the missed dose estimate for unbadged individuals working in 

radiologically controlled areas would be one-half of 10% of 3 rem/qtr or 600 mrem/yr. A lognormal 

distribution should be assumed, with the upper 95% dose estimate for these individuals therefore 

estimated to be 1.2 rem (NIOSH 2006). 

In Production Areas

An RFP External Dosimetry coworker study has been performed and is documented in OTIB-0058 

(ORAU 2006d). The coworker study document addresses the application of these results to the dose 

reconstructions as appropriate. 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 60 of 90

Table 6-31. 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

Pa-231 1.04E-13 3.84E-03

Pa-234a 5.44E-10 2.01E+01

Pa-234m 3.40E-07 1.26E+04

Pb-210a 7.10E-19 2.63E-08

Pb-211 1.28E-15 4.75E-05

Pb-214a 7.00E-17 2.59E-06

Po-210 2.16E-19 7.99E-09

Po-211 3.50E-18 1.30E-07

Po-214 6.99E-17 2.59E-06

Po-215 1.28E-15 4.75E-05

Po-218 7.00E-17 2.59E-06

Ra-223 1.28E-15 4.75E-05

Ra-226 7.21E-17 2.67E-06

Rn-219 1.28E-15 4.75E-05

Rn-222 7.00E-17 2.59E-06

Th-227 1.39E-15 5.15E-05

Th-230 3.33E-13 1.23E-02

Th-231a 4.90E-09 1.81E+02

Th-234a 3.40E-07 1.26E+04

Tl-207 1.28E-15 4.74E-05

U-234 3.70E-08 1.37E+03

U-235 4.90E-09 1.81E+02

U-238 3.40E-07 1.26E+04

a. Significant progeny (included in VARSKIN 

Mod 3). 

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 2006c)]. 

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 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 61 of 90

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.

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 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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 62 of 90

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.

[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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 63 of 90

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.

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 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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 64 of 90

[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.

[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. 01 Effective Date: 02/08/2007 Page 65 of 90

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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 seiverts. 

(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. 01 Effective Date: 02/08/2007 Page 70 of 90

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. 01 Effective Date: 02/08/2007 Page 71 of 90

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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 72 of 90

shallow absorbed dose (Ds) 

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

composition.

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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 73 of 90

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

74

A-2 Occupational Dose Report reviewed 6-4-02, page 2 .............................................................. 75

A-3 Dosimetry History by Individual dated 3-10-03, page 1 .......................................................... 76

A-4 Dosimetry History by Individual dated 3-10-03, page 2 .......................................................... 77

A-5 Dosimetry History by Individual dated 3-10-03, page 3 .......................................................... 78

A-6 Health Physics Yearly External Exposure Activity Run, 1953 to 1958 .................................... 79

A-7 Health Physics Yearly External Exposure Run, 1959 to 1963 ................................................ 80

A-8 Health Physics Yearly External Exposure Activity Run, 1964................................................. 81

A-9 External Dosimetry Detail, 1981 to 1983 ................................................................................ 82

A-10 Health Physics External Radiation Exposure Report, 1967 .................................................... 83

A-11 Radiation Dosimetry Individual Lifetime Report dated 6-4-02................................................. 84

A-12 Radiation Health Records System – TLD Data....................................................................... 85

A-13 Radiation Dosimetry Termination Report dated 9-17-96, page 1............................................ 86

A-14 Radiation Dosimetry Termination Report dated 9-17-96, page 2............................................ 87

A-15 Occupational Radiation Exposure Information, 1988.............................................................. 88
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 74 of 90

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. 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 76 of 90

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. 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 78 of 90

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. 01 Effective Date: 02/08/2007 Page 80 of 90

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 8 of 16

Figure A-7. Health Physics Yearly External Exposure Run, 1959 to 1963. 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 81 of 90

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. 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 85 of 90

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 Termination Report dated 9-17-96, page 1. 
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 87 of 90

ATTACHMENT A

EXAMPLE EXTERNAL DOSIMETRY RECORD DOCUMENTS

Page 15 of 16

Figure A-14. Radiation Dosimetry Termination Report dated 9-17-96, page 2. 
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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. 01 Effective Date: 02/08/2007 Page 89 of 90

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.
Document No. ORAUT-TKBS-0011-6 Revision No. 01 Effective Date: 02/08/2007 Page 90 of 90

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.