Page 1 of 61

DOE Review Release 12/19/2017 

Pinellas Plant – Occupational External Dose

ORAUT-TKBS-0029-6 Rev. 02

Effective Date: 12/11/2017

Supersedes: Revision 01

Subject Expert(s): Brian P. Gleckler and Peter A. Darnell (NIOSH)

Document Owner 

Approval: Signature on File

Brian P. Gleckler, Document Owner

Approval Date: 12/11/2017

Concurrence: Signature on File

John M. Byrne, Objective 1 Manager

Concurrence Date: 12/08/2017

Concurrence: Signature on File

Scott R. Siebert, Objective 3 Manager

Concurrence Date: 12/11/2017

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 12/08/2017

Approval: Signature on File

James W. Neton, Associate Director for Science

Approval Date: 12/11/2017

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

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

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0029-6 Revision No. 02 Effective Date: 12/11/2017 Page 2 of 61

 

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

09/15/2005 00 New Technical Basis Document for Pinellas – Site External Dose. 

First approved issue. Training required: As determined by the Task 

Manager. Initiated by Mark D. Notich.

08/03/2006 00 PC-1 Approved page change revision to incorporate changes according to 

the public outreach meetings in 09/2004 and 11/2005 and claimant 

communications. Updated required language on pages 6 and 7 in 

Section 6.1. Added Purpose and Scope Sections on page 7. 

Changes occurred on page 10, in Section 6.1.4, page 27 in Section 

6.4.2, and on pages 29 - 32 in the Reference Section. As a result of 

internal formal review, changes were made to Tables 6-4, 6-5, 6-11, 

and 6-12 on pages 16, 17, 25, and 27, respectively in Sections 6.2.3 

and 6.4.1.1 and 6.4.2. Wording was added or changed on pages 10, 

15, 19, 23, 24, and 27, in Sections 6.1.1, 6.2.3, 6.2.4, 6.3, and 

6.4.2.1. No sections were deleted. This revision results in a 

reduction in assigned dose and no PER is required. Training 

required: As determined by the Task Manager. Initiated by Paul J. 

Demopoulos. Approval:

Signature on File 07/25/2006

Paul J. Demopoulos, TBD Team Leader

Signature on File 07/25/2006

John M. Byrne, Task 3 Manager

Signature on File 07/24/2006

Edward F. Maher, Task 5 Manager

Signature on File 07/27/2006

Kate Kimpan, Projector Director

Signature on File 08/03/2006

James W. Neton, Associate Director for Science

11/08/2006 00 PC-2 Approved page change revision to incorporate changes according to 

Task 5 comment and Task 7 amendment to page 27 and 28 in 

Section 6.4.1.1. Corrects table numbering on page 28 and 29. Adds 

two references to page 34. Corrects references on pages 20, 24, 

and 32. No sections were deleted. Incorporates internal formal 

review comments. Incorporates NIOSH formal review comments. 

Minor editorial changes were made on pages 24 and 29. Deletes 

Task 7 amendment to page 27 and 28 in Section 6.4.1.1. Corrects 

Table 6-5 rows 2 and 3 on page 18. This revision results in an 
Document No. ORAUT-TKBS-0029-6 Revision No. 02 Effective Date: 12/11/2017 Page 3 of 61

 

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

11/08/2006 

(continued)

00 PC-2 

(continued

increase in assigned dose and a PER is required. Training required: 

As determined by the Task Manager. Initiated by Paul J. 

Demopoulos. Approval:

Signature on File 10/19/2006

Paul J. Demopoulos, Document Owner

Signature on File 10/20/2006

John M. Byrne, Task 3 Manager

Signature on File 10/24/2006

Edward F. Maher, Task 5 Manager

Signature on File 10/23/2006

Kate Kimpan, Projector Director

Brant A. Ulsh Signature on File for 11/08/2006

James W. Neton, Associate Director for Science

04/28/2011 01 This technical basis document was predominantly revised to address 

SC&A’s issues with the document, as identified in SCA-TR-TASK1-

0015. Adjustments to certain photon dosimeter results now need to 

be made per Section 6.3.1.1. Photon LOD values specific to the 

Mound dosimeters are now provided in Table 6-10. Significant 

changes were also made on how neutron doses are assessed, and 

the new approaches are described in Section 6.3.2. The applicability 

dates for the various types of neutron dosimeters used at the site 

were changed, and are now in Table 6-11. In addition, a number of 

editorial changes were made to this document. Incorporates formal 

internal and NIOSH review comments. Constitutes a total rewrite of 

the document. Training required: As determined by the Objective 

Manager. Initiated by Brian P. Gleckler.

12/11/2017 02 Revision initiated to address update the missed dose tables in 

Section 6.4 to provide consistent LOD data for dose reconstruction. 

In addition, the following changes were also made: editorial changes, 

information updated, redundant information eliminated, and several 

sections were reorganized. Incorporates formal internal and NIOSH 

review comments. Constitutes a total rewrite of the document. 

Training required: As determined by the Objective Manager. 

Initiated by Brian P. Gleckler.

TRADEMARK INFORMATION

Radiflo® is a registered trademark of Isovac Engineering in the United States and/or other countries.

Landauer® is a registered trademark of R. S. Landauer, Jr. & Company in the United States and/or 

other countries.

Lexan® is a registered trademark of Sabic Global Technologies in the United States and/or other 

countries.
Document No. ORAUT-TKBS-0029-6 Revision No. 02 Effective Date: 12/11/2017 Page 4 of 61

 

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ............................................................................................................... 6

6.1 Introduction .............................................................................................................................. 8

6.1.1 Purpose........................................................................................................................ 9

6.1.2 Scope........................................................................................................................... 9

6.1.3 Overview ...................................................................................................................... 9

6.2 Dose Reconstruction Parameters........................................................................................... 10

6.2.1 Radiation Sources ...................................................................................................... 10

6.2.1.1 Radioactive Materials................................................................................... 10

6.2.1.2 Radiation Generating Devices...................................................................... 11

6.2.2 Workplace Radiation Fields ........................................................................................ 12

6.2.2.1 Photon and Electron Radiation Fields .......................................................... 12

6.2.2.2 Neutron Radiation Fields.............................................................................. 13

6.2.3 Dosimetry Technology................................................................................................ 15

6.2.4 Dosimetry Performance and Calibration ..................................................................... 22

6.2.4.1 Performance Testing.................................................................................... 22

6.2.4.2 Calibration.................................................................................................... 23

6.2.5 Historical Administrative and Reporting Practices....................................................... 24

6.2.5.1 Administrative Practices............................................................................... 24

6.2.5.2 Reporting Practices...................................................................................... 26

6.3 Adjustments to Recorded Dose.............................................................................................. 26

6.3.1 Photon Dose Adjustments .......................................................................................... 27

6.3.1.1 Signal Fading Adjustments........................................................................... 27

6.3.2 Neutron Dose Adjustments......................................................................................... 27

6.3.2.1 Radiation Weighting Factor Adjustments...................................................... 27

6.3.2.2 Track Fading Adjustments............................................................................ 28

6.3.2.3 Signal Fading Adjustments........................................................................... 28

6.3.2.4 Neutron Energy Response Adjustments....................................................... 29

6.3.3 Electron Dose Adjustments ........................................................................................ 30

6.4 Missed Dose .......................................................................................................................... 30

6.5 Unmonitored Dose ................................................................................................................. 33

6.5.1 Unmonitored Workers................................................................................................. 33

6.5.2 Workers Whose Dosimetry Records Are Missing........................................................ 34

6.6 Uncertainty............................................................................................................................. 34

6.6.1 Beta-Gamma Dosimeter Uncertainty .......................................................................... 34

6.6.2 Neutron Dosimeter Uncertainty................................................................................... 35

6.7 Attributions and Annotations .................................................................................................. 35

References ........................................................................................................................................ 36

Glossary ............................................................................................................................................ 44
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ATTACHMENT A EXAMPLES OF PINELLAS PLANT EXTERNAL DOSIMETRY RECORDS .... 50

ATTACHMENT B BASIS FOR UNMONITORED EXTERNAL DOSE ASSIGNMENT .................. 55

LIST OF TABLES

TABLE TITLE PAGE

6-1 Electron and photon radiation energies and percentages....................................................... 13

6-2 Neutron radiation energies and percentages.......................................................................... 15

6-3 Pinellas Plant historical dosimetry events............................................................................... 17

6-4 Whole body beta-gamma dosimeters used at Pinellas Plant .................................................. 19

6-5 Whole body neutron and beta-gamma-neutron dosimeters used at Pinellas Plant ................. 20

6-6 Wrist beta-gamma dosimeters used at Pinellas Plant............................................................. 21

6-7 Ring beta-gamma dosimeters used at Pinellas Plant ............................................................. 22

6-8 Annual whole body radiation exposure information for the Pinellas Plant............................... 25

6-9 Neutron radiation energies and ICRP 60 correction factors.................................................... 28

6-10 Summary of neutron dosimeter correction factors for neutron generator operations 

(1957–1997)...........................................................................................................................29

6-11 Summary of neutron dosimeter correction factors for RTG operations (1975–1991) .............. 30

6-12 Whole-body beta-gamma dosimetry missed doses for Neutron Generator Operations 

(1957–1997)...........................................................................................................................31

6-13 Whole-body beta-gamma dosimetry missed doses for RTG Operations (1975–1991) ........... 31

6-14 Whole-body neutron dosimetry missed doses for Neutron Generator Operations (1957–

1997)......................................................................................................................................32

6-15 Whole-body neutron dosimetry missed doses for RTG Operations (1975–1991) ................... 32

6-16 Wrist beta-gamma dosimetry missed doses........................................................................... 33

6-17 Finger ring beta-gamma dosimetry missed doses .................................................................. 33

B-1 Evaluation of Pinellas Plant doses available in the SRDB...................................................... 57

B-2 Evaluation of POCs associated with assigning unmonitored doses as either photon or 

neutron dose..........................................................................................................................61

LIST OF FIGURES

FIGURE TITLE PAGE

6-1 Neutron energy spectrum for a D-T reaction .......................................................................... 14

6-2 Neutron energy spectrum for a D-D reaction.......................................................................... 14

6-3 RTG plutonium source neutron energy spectrum – PuO2 microspheres................................. 15

6-4 Redacted example of a Landauer dosimetry report from 1978 ............................................... 18

A-1 Example of dosimetry records used during 1957–1959.......................................................... 51

A-2 Example of dosimetry records used during 1959–1965.......................................................... 52

A-3 Example of dosimetry records used during 1964–1970.......................................................... 52

A-4 Example of dosimetry records used during 1971–1980.......................................................... 53

A-5 Example of dosimetry records used during 1980–1987.......................................................... 54

A-6 Example of dosimetry records used during 1988–1997.......................................................... 54
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ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission

ALARA as low as reasonably achievable

AWE atomic weapons employer

Bq becquerel

CFR Code of Federal Regulations

Ci curie

cm centimeter

D deuterium (2

H)

DCAS Division of Compensation Analysis and Support

DCF dose conversion factor 

DOE U.S. Department of Energy

DOL U.S. Department of Labor

DOELAP U.S. Department of Energy Laboratory Accreditation Program

DU depleted uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ft foot

g gram

GE General Electric Company

GEND GE Neutron Devices 

GENDD GE Neutron Devices Department 

GEPP GE Pinellas Plant

GEXF GE X-Ray Division in Florida

GEXM GE X-Ray Division in Milwaukee, Wisconsin

hr hour

HSR Health and Safety Record (system)

ICRP International Commission on Radiological Protection

in. inch

IREP Interactive RadioEpidemiological Program

keV kiloelectron-volt, 1,000 electron-volts

Landauer R. S. Landauer Jr. & Co.

LOD limit of detection

MeV megaelectron-volt, 1 million electron-volts

mR milliroentgen

mrem millirem

n neutron

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH-OCAS Claims Tracking System

NDD Neutron Devices Department

NTA nuclear track emulsion, type A
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ORAU Oak Ridge Associated Universities

ORAUT Oak Ridge Associated Universities Team

OW open window

PER Program Evaluation Report

PAO Pinellas Area Office

POC probability of causation

RBE relative biological effectiveness

RTG radioisotopically-powered thermoelectric generator

SEC Special Exposure Cohort

SNL Sandia National Laboratories

SRDB Site Research Database

SRDB Ref ID Site Research Database Reference Identification (number)

T tritium (3

H)

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

yr year

μCi microcurie

α alpha particle

§ section or sections
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6.1 INTRODUCTION

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

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

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

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

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

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

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

of the individual work required for each dose reconstruction.

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

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

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

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

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

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

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

§ 7384l(5). On the other hand, a DOE facility is defined as “any building, structure, or premise, 

including the grounds upon which such building, structure, or premise is located ... in which 

operations are, or have been, conducted by, or on behalf of, the [DOE] (except for buildings, 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

dose reconstructions for covered employees (NIOSH 2010):

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

conventional structures

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

reasons
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6.1.1 Purpose

The purpose of this technical basis document (TBD) is to document the external dosimetry program 

and practices at the Pinellas Plant, and to provide the technical basis to be used to evaluate the 

external occupational radiation dose for EEOICPA claims.

6.1.2 Scope

This TBD provides supporting documentation to assist in the evaluation of occupational external 

doses in accordance with the guidelines described in External Dose Reconstruction Implementation 

Guideline (NIOSH 2007). NIOSH considers the available data and methods for performing external 

dose reconstruction to be adequate for estimating with sufficient accuracy the external doses at the 

Pinellas Plant throughout its entire history.

This TBD describes the external dosimetry program at the Pinellas Plant. It discusses dose 

reconstruction, practices and policies at the Pinellas Plant, and dosimeter types and technologies for 

measuring dose from the different types of radiation in the work environment. It also discusses the 

specific details of the evaluation of doses that were measured from exposures to electron (beta 

particles), photon (X-rays and gamma rays), and neutron radiation; sources of bias; workplace 

radiation field characteristics; responses of different dosimeters in the workplace radiation fields; and 

adjustments to the recorded dose measured by these dosimeters during specific years.

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

justification, or clarification of the associated information, are presented in Section 6.7.

6.1.3 Overview

This TBD is Part 6 of the Pinellas Plant Site Profile. A site profile provides a summary of information 

about a site that is relevant to the dose reconstruction process.

The Pinellas Plant has been known by several names throughout its history. Those names include: 

908 Plant, Pinellas Peninsula Plant, General Electric (GE) X-ray Division-Florida (GEXF), GE Neutron 

Devices Department (GENDD), GE Neutron Devices (GEND), GE Pinellas Plant (GEPP), and the 

Pinellas Plant. This document uses the latter for convenience.

The General Electric Company built and operated the Pinellas Plant for DOE from its initial startup in 

January 1957 until June 1992. In June 1992, Martin Marietta Specialty Components (MMSC) took 

over as the managing and operating contractor for the Pinellas Plant. In 1994, Lockheed merged with 

Martin Marietta and the managing and operating contractor for the Pinellas Plant was renamed 

Lockheed Martin Specialty Components (LMSC). The Pinellas Plant completed its war reserve 

fabrication of neutron generators at the end of September 1994, and began the transition from a 

defense mission to an environmental management mission. That transition included a number of 

decontamination and decommissioning activities that allowed the Plant to be turned over for 

commercial uses. LMSC continued as the managing and operating contractor until decontamination 

and decommissioning activities ended in 1997 (ORAUT 2011b).

The Plant was built to manufacture neutron generators, a principal component in nuclear weapons. 

The neutron generators consisted of a miniaturized linear ion accelerator with pulsed electric power 

supplies. The ion accelerator, or neutron tube, required ultraclean, high-vacuum technology; hermetic 

seals between glass, ceramic, glass-ceramic, and metal materials; and high-voltage generation and 

measurement technology. The Plant manufactured only neutron generators for its first 10 years of 

operation. It later manufactured other products including neutron detectors, radioisotopically-powered 

thermoelectric generators (RTGs), high-vacuum switch tubes, specialty capacitors, and specialty 
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batteries (Weaver 1990). As part of its program to promote commercial uses of the site, DOE sold 

most of the Plant to the Pinellas County Industry Council in March 1995 and leased back a portion 

through September 1997 to complete safe shutdown and transition activities (MMSC 1996).

6.2 DOSE RECONSTRUCTION PARAMETERS

6.2.1 Radiation Sources

The manufacture of Pinellas Plant products required the use of radioactive materials and radiation 

generating devices. A variety of radioactive materials and radiation generating devices were used at 

the Pinellas Plant. The following sub-sections identify the major sources of radioactivity and/or 

radiation at the plant. 

6.2.1.1 Radioactive Materials

The predominant radioactive materials used at the Plant included tritium, 85Kr, depleted uranium (DU), 

and plutonium. A wide variety of other radionuclides were used at the Pinellas Plant; however, the 

uses of these radionuclides were mostly limited to sealed and plated check sources, static meter 

sources, explosive meter sources, heat sources, calibration sources, thickness gauges, gas 

chromatograph sources, dew point measurement sources, and static eliminator sources (Author 

unknown undated b). 

The predominant source of electron radiation at the Pinellas Plant was tritium, but this radionuclide 

emits only a low-energy beta particle with an average energy of 5.7 keV. Because electrons below 

15 keV do not have sufficient energy to penetrate the epidermal layer of the skin (NIOSH 2007), 

tritium is not considered to be an external radiation hazard. 

Krypton-85, a beta and gamma emitter, was used in two leak detection systems (Radiflo and 

TRACER-flo systems) as part of the Pinellas Quality Control Program. The leak detection systems 

were housed in separate rooms and surrounded by ventilation shrouds. Each shroud was connected 

to ductwork that exhausted to the east main exhaust stack. Because it is a noble gas, 85Kr can deliver 

a whole-body dose with the possibility of a significant skin dose from electron radiation. Because of 

engineering controls such as shielding and ventilation, significant external exposures to 85Kr gas at 

the Pinellas Plant were unlikely, but did occur during some accidents (MMSC 1993, Author unknown, 

undated d, ORAU 2007). The 85Kr was first introduced into one of the leak detection systems on 

September 24, 1963 (GE 1957–1967, p. 253). By the end of 1996, all 85Kr had been removed from 

the site (LMSC 1997, p. 12). 

Beginning in the late-1970’s, portable radiation dose rate instruments were calibrated in Building 800 

using a Model 81-12 Beam Calibrator manufactured by J. L. Shepard and Associates (GE 1977, 

Author unknown undated e). The Model 81-12 Beam Calibrator contained a 120-Ci 137Cs source, 

which was in the form of a sealed source (GE 1977, Author unknown undated e). The 137Cs was in a 

sealed source in a shielded cabinet in the concrete Building 800, so the probability of a worker 

receiving electron radiation exposures from 137Cs was remote. 

In either 1965 or 1968, the Pinellas Plant started using depleted uranium (DU), consisting mainly of 

238U, for its tritium storage beds (Phillips 1975, GE 1979). According to one document, the tritium 

storage beds using DU were 14 years old in 1979, indicating that they were first used in 1965 (GE 

1979). Whereas, another document stated that the tritium storage beds using DU were 7 years old in 

1975, indicating that they were first used in 1968 (Phillips 1975). In the new tritium storage beds, DU 

metal was used to store tritium as uranium tritide (UT3). The DU metal in the tritium storage beds was 

sealed in stainless steel canisters (GE 1979; Ward 1973). The DU inside the tritium storage beds 
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presents no significant external radiation hazard, due to the relatively small quantities used, low 

specific activity for DU, and its non-penetrating radiation.

The first plutonium that was received at the Pinellas Plant was a 7 g Pu source (predominately 239Pu), 

which was received in September 1957. The source was an encapsulated plutonium-beryllium (PuBe) neutron source that was used for calibrating health physics monitoring equipment (Author 

unknown undated b, GE 1957–1973). In December 1978, this Pu-Be neutron source was shipped 

offsite for disposal (Author unknown undated c, Various authors 1961–1990). In August 1979, the 7 g 

Pu-Be neutron source was replaced with a Pu-Be source containing 32 g of Pu, which was also 

predominately 239Pu (Author unknown undated c, Various authors 1961–1990). In October 1990, the 

32 g Pu-Be neutron source was shipped offsite for disposal (Author unknown undated c, Various 

authors 1961–1990). 

The thimble size triply encapsulated plutonium oxide (238PuO2) heat sources that were used for the 

radioisotopically-powered thermoelectric generators (RTGs) did not start arriving at the Pinellas Plant 

until November 1975 (Author unknown undated b, GE 1982). In November 1975, the site received 

seven plutonium heat sources (Author unknown undated b). The radionuclide composition of each 

sealed plutonium heat source was approximately 80% 238Pu, 16% 239Pu, 3% 240Pu, and 1% of other 

radionuclides by mass (GE 1982). There were two different types of plutonium heat sources, 8.75 g 

sources and 10 g sources (GE 1982). Between 1975 and 1991, monthly inventories of the RTG heat 

sources ranged from 31.2 g to 4,397.5 g of Pu (Author unknown undated c). The Pinellas Plant also 

had several small plutonium sources that were as alpha check sources for checking instruments 

(Author unknown undated b), but those were insignificant from an external exposure perspective. All 

plutonium, with the exception of calorimeter sources and very small instrument calibration check 

sources, was removed from the Plant by February 1991 (Author unknown undated c, MMSC 1992). 

In approximately 1988, the Pinellas Plant acquired a 10 Ci 241Am source for unknown reasons. The 

source was an encapsulated americium-beryllium (Am-Be) neutron source, which was likely used as a 

calibration source. By October 1991, the Am-Be neutron source was inactive and in the process of 

being excessed (Author unknown undated b). 

6.2.1.2 Radiation Generating Devices

Radioactive materials were not the only source of ionizing radiation at the Pinellas Plant. Many 

pieces of equipment produced and used at the Plant were capable of generating radiation. Unlike 

radioactive materials, radiation-generating devices emit radiation only when they are connected to a 

power supply and are activated. When not in use, these devices did not emit radiation.

The most common type of radiation-generating device at the Pinellas Plant was its primary product, 

the neutron generator. Neutron generators are miniaturized linear ion accelerators (DOE 1987). A 

pulsed electric power supply accelerates deuterons (i.e., deuterium nuclei) into either a tritium or 

deuterium target, depending on the type of neutron generator, to create a controlled source of 

neutrons (DOE 1987; NCRP 1983; Weaver 1994a). The neutrons are generated by either a 

T(d,n)4

He or D(d,n)3He fusion reaction. Most units produced at the Pinellas Plant were the T(d,n)4He 

type that produced 14-MeV neutrons; however, a few were constructed to produce 2.5-MeV neutrons 

from the D(d,n)3

He reaction (Weaver 1994a; NCRP 1983). The neutron generators also produce 

some X-rays by other interactions within the accelerator (NCRP 1983). This radiation was generated 

by testing either the neutron tubes (a component in the neutron generators) or the completed neutron 

generators. 

An ion accelerator, a Model 200 HP Ion Implanter that was manufactured by Accelerator Inc., was 

also used at the Pinellas Plant (Malbrough 1983). It was a Cockroft-Walton-type linear ion accelerator 

and was first installed in 1975 in Area 161 of Building 100 for use by the Chemistry Laboratory (GE 
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1977a, p. 23; Malbrough 1983). The accelerator was originally used for ion implantation work and 

eventually for target assessment work prior to being relocated (GE 1977a; Malbrough 1983). In 1979, 

the accelerator was relocated to Building 800. After the accelerator’s relocation, it was used for a 

larger variety of activities that included target assessment; material analysis; low-energy nuclear, 

solid-state, and atomic physics; and material science (Malbrough 1983). Personnel working with this 

accelerator were required to wear dosimeters that measured both photon and neutron doses (Weaver 

1994a).

Table 2-1 of the Pinellas Plant – Site Description (ORAUT 2011b) provides a listing of the other

radiation generating devices and their locations at the Pinellas Plant. These devices were not a 

radiation concern under normal operating conditions, because of engineering controls (e.g. shielding, 

access controls, system interlocks, etc...). However, there were recorded incidents at the Pinellas 

Plant involving these types of devices that resulted in external radiation exposures to a few of its 

workers. The listing of unusual events and incidents provided in Table 2-4 of the Pinellas Plant – Site 

Description (ORAUT 2011b) includes several radiation exposure incidents involving radiation 

generating devices. Based on a review of the available claim records, the Pinellas Plant appears to 

have assessed any potential doses from those incidents and included them in the workers’ dosimetry 

records. 

6.2.2 Workplace Radiation Fields

Potential sources for workplace radiation fields at Pinellas can be placed in two categories, 

radionuclide sources and machine-generated sources. The only open-area radiation fields to be 

routinely encountered by workers were from the testing of neutron tubes and neutron generators, the 

use of machine-generated X-rays, and areas where work was performed with the RTG heat sources. 

During some accidents involving either of the two leak detection systems (Radiflo and TRACER-flo

systems), open-area radiation fields were temporarily present in the rooms that housed those systems 

due to 85Kr gas leaks (MMSC 1993, Author unknown, undated d, ORAU 2007). During those 

incidents, non-routine external exposures to 85Kr gas occurred, but those exposures were limited to 

personnel working in the rooms that housed the two leak detection systems due to the engineering 

controls for those systems. 

6.2.2.1 Photon and Electron Radiation Fields 

The NIOSH Interactive RadioEpidemiological Program (IREP) uses three photon energy groups 

(below 30 keV, 30 to 250 keV, and above 250 keV) (NIOSH 2007). For electron exposures, there are 

two electron energy groups (equal to or below 15 keV and above 15 keV); however, only the above 

15 keV group is applicable to external dose. 

The majority of the photon radiation exposures in the neutron generator production areas were from 

the testing of neutron tubes and neutron generators. In the RTG production areas (Building 400), the 

majority of the photon radiation exposures were from the plutonium oxide (238PuO2) heat sources. 

The only potentially significant sources of electron radiation with sufficient energy to penetrate the skin 

are 85Kr that is used with the two leak detection systems (Radiflo and TRACER-flo systems). Electron 

radiation exposures were also possible for X-ray diffraction and electron beam devices if containment 

of the beams was compromised. However, it was more probable that any exposures from these 

devices would have been from X-rays or bremsstrahlung production and not from a free electron 

beam. The exposures, if diffuse, would have been monitored by film badge or TLD. 

Table 6-1 lists electron and photon energies and percentages for the various locations at the Pinellas 

Plant. For most locations, specific photon energy distribution information was not available. 

Therefore, with the exception of one location, 100% of the photons were assumed to be 30 to 250 keV 
Document No. ORAUT-TKBS-0029-6 Revision No. 02 Effective Date: 12/11/2017 Page 13 of 61

 

photons, which is the most favorable photon energy interval. For exposures to 85Kr gas in the area 

housing the two leak detection systems (Radiflo and TRACER-flo systems), a different photon energy 

distribution was used, based on the X-rays and gamma-rays being emitted from 85Kr. 

Table 6-1. Electron and photon radiation energies and percentages. 

Location Process type

Radiation type, energy 

group, and percentage

Buildings 100, 200, 300 Neutron generator production areas Photon, 30–250 keV, 100%

Building 100, Area 109a Leak detection systems area with 85Kr

(Radiflo and TRACER-flo systems area) 

Electron, >15 keV, 100%; 

Photon 30–250 keV, 86%; 

Photon, >250 keV, 14%

Building 400 RTG production Photon, 30–250 keV, 100%

Building 800b Calibration and ion accelerator Electron, >15 keV, 100%; 

Photon, 30–250 keV, 100%

a. The location of the two leak detection systems may have changed during their use at the Pinellas Plant. Because 

exposure to 85Kr gas only occurred during accidents with these leak detection systems, the parameters for the 

leak detection systems area are only applicable to accidental exposures to 85Kr gas. 

b. Electron doses would not normally be assessed for this location. The assessment of electron dose for Building 

800 would only be performed for workers accidentally exposed to the 120 Ci 137Cs calibration source, if any such 

exposure ever occurred. 

6.2.2.2 Neutron Radiation Fields

The NIOSH Interactive RadioEpidemiological Program (IREP) uses five neutron energy groups (below

10 keV, 10 to 100 keV, 0.1 to 2.0 MeV, 2.0 to 20.0 MeV, and above 20.0 MeV) (NIOSH 2007). 

There were two distinct sources of neutrons at the Pinellas Plant. The first distinct sources of 

neutrons, were the neutrons produced by neutron generator sources, such as those generated by the 

testing of neutron tubes, testing of neutron generators, and operation of ion accelerator in 

Building 800. Those sources generated neutrons by either a T(d,n)4

He or D(d,n)3He fusion reaction 

(also notated as D-T and D-D reactions), and only produce neutrons when electrically activated. Most 

of these sources produced 14-MeV neutrons from the T(d,n)4

He reaction; however, a few produced 

2.5-MeV neutrons from the D(d,n)3

He reaction (Weaver 1994a; NCRP 1983). Note that these energy 

values are approximate and vary slightly from one reference to the next. This is mostly due to the fact 

that these sources actually produce a distribution of neutrons at various energies versus neutrons at a 

single discrete energy. This is illustrated in Figures 6-1 and 6-2, which are neutron spectrums for 

each of these reactions (Ing, Cross, and Tymons 1977). 

The second distinct source of neutrons, were the sealed PuO2 heat sources used for the RTGs. The 

RTG PuO2 heat source neutron spectrum is illustrated in Figure 6-3 (Greene 1984b). The spectrum 

was probably from the Mound Laboratory and used to analyze the effectiveness of various Landauer 

Neutrak TLDs by the Pinellas Health Physics Department (Burkhart 1987a). It was determined that 

the Landauer dosimeters responded to only about 67% of the dose equivalent for the RTG PuO2 heat 

source spectrum. 

Table 6-2 lists the IREP neutron energy groups and fractions (as percentages) that are applicable to 

the various locations at the Pinellas Plant. 
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Figure 6-1. Neutron energy spectrum for a D-T reaction (Ing, 

Cross, and Tymons 1977). 

Figure 6-2. Neutron energy spectrum for a D-D reaction (Ing, 

Cross, and Tymons 1977). 
Document No. ORAUT-TKBS-0029-6 Revision No. 02 Effective Date: 12/11/2017 Page 15 of 61

Figure 6-3. RTG plutonium source neutron energy spectrum – 

PuO2 microspheres (Greene 1984b). 

Table 6-2. Neutron radiation energies and percentages. 

Energy groups and 

Location Process type percentage

Building 100 Neutron generator production areas Neutron, 2–20 MeV, 100%

Building 400 RTG production Neutron, 0.1–2 MeV, 50%

Neutron, 2–20 MeV, 50%

Building 800b Ion accelerator Neutron, 2–20 MeV, 100%

6.2.3 Dosimetry Technology

For the period of 1957–June 1974, GEND-Health Physics conducted radiation dosimetry 

management and analysis through in-plant processing of X-ray and neutron-sensitive photographic 

films (Burkhart 1987b). For that period, no information regarding the specific design parameters for

the film dosimeters used at the Pinellas Plant (e.g. film type, filters used, dosimeter holder design, 

etc...) could be found. During the period that the Pinellas Plant read its own dosimeters (1957–June 

1974), the available film work sheets indicate that only the open window (OW) film was read from the 
Document No. ORAUT-TKBS-0029-6 Revision No. 02 Effective Date: 12/11/2017 Page 16 of 61

film dosimeters and the OW result was reported as photon dose (GE 1957–1990, pages 67 and 99). 

Because the OW portions of the early film dosimeters typically over-responded to photons with 

energies <500 keV (ORAUT 2005c, 2006) and because a significant portion of the Pinellas Plant’s 

photon doses were attributable to low energy photons, this practice likely resulted in the reported 

photon doses for the period of 1957–June 1974 being overestimates of the workers actual photon 

doses. Because specific information on dosimeter design and dosimeter calibration could not be 

confirmed for the period of 1957– June 1974, no adjustments to the reported photon doses are 

recommended, to account for the over-response that the dosimeters likely had to the photon radiation 

at the Plant. The practice of only reading the OW film and reporting the result as photon dose would 

also explain why non-penetrating doses such as electron doses were not reported during this era. In 

addition, any potential electron doses received by the workers would have been accounted for and 

reported as photon doses by this practice.

An unsigned note in a collection of highly varied examples of dosimetry records, which include a 

number of records from GEXM, indicates that personnel monitoring for neutrons might not have 

begun at the Pinellas Plant until 1960, and neutron doses for the period from 1957 through May 1960 

might have been estimated from area monitors (GE 1957–1990). Even if the note is accurate about 

neutron monitoring at the Plant, it should not have an adverse effect on the neutron dose calculations 

for that period. Because the same whole-body dosimeters were used for personnel and area 

monitoring, the neutron dose calculations would be the same regardless of whether the worker’s 

neutron doses were based on a personnel dosimeter or a dosimeter used to monitor an area.

Beginning in July 1974, the Plant began using dosimetry provided exclusively by Landauer. The 

original Landauer dosimetry was based on film badge technology (Ward 1974). Landauer processed 

the dosimetry and provided exposure reports to GEND for review after badge processing. Figure 6-4 

is an example of a Landauer dosimetry report from 1978, with all personal information redacted (GE 

1974–1980). Information in the Landauer reports included personnel data (identification number, 

name, social security number, and birth date), dosimeter type, deep and shallow exposure for the 

reporting period (i.e., monthly), and cumulative totals for deep and shallow exposures for the calendar 

quarter, year to date, and permanent exposure (lifetime exposure at the Pinellas Plant). The 

exposure information was entered in each person’s exposure history by hand or into a Pinellas-based 

computer system and GE’s Corporate Health and Safety Record (HSR) system (Richards 1986). The 

Landauer records are not provided by the DOE for the EEOICPA claims, examples of those records 

are provided in Attachment A.

Starting in mid-1978, the Plant began using the Landauer polycarbonate plastic dosimeter for 14-MeV 

neutrons, and continued to use photographic film processing for 2.5-MeV neutrons, X-ray, beta, and 

gamma exposures, and Landauer thermoluminescent dosimeter (TLD) rings for hand monitoring 

(Burkhart 1987b).

From October 1979 through September 1987, the Plant used dosimeters from Mound Laboratory for 

evaluating exposures to X-rays and 2-MeV average neutrons from the handling of sealed 238PuO2

sources during the production of RTG units in Building 400, and continued to use the Landauer 

dosimetry discussed above for exposures from all other radiation sources, including 14-MeV neutrons. 

Problems related to the DOE Laboratory Accreditation Program (DOELAP) testing with the Mound 

dosimetry and the equivalent performance of Landauer neutron dosimetry led GEND to discontinue 

use of the Mound dosimetry in October 1987 (Burkhart 1987b). Beginning in 1990, earlier dosimetry 

technology was replaced with Landauer TLD dosimetry that was used until the end of nuclear 

development and testing operations in 1994. Table 6-3 summarizes the significant historical events 

for the dosimetry program at the Pinellas Plant.
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Table 6-3. Pinellas Plant historical dosimetry events. 

Date Event Reference

1957 In-plant processing of gamma and neutron film dosimeters initiated. Burkhart 1987a

April 1957 New employee orientation in radiation safety offered. Burkhart 1990

October 1957 Measured neutron dose rates at all test positions. Burkhart 1990

November 1957 Measured neutron output of neutron generators, and was 

determined to be equivalent to 10 mrem/pulse at 1 in.

Burkhart 1990

December 1957 SNL asked to provide film badges. However, there is no indication 

that SNL ever provided any film badges for the Pinellas Plant.

Burkhart 1990

January 1960 Full-time Health Physics representative assigned to Area 108. Burkhart 1990

November 1963 Began use of wrist badges in place of ring badges for limited 

number of employees.

Burkhart 1990

February 1965 GEXM memorandum comparing performance of two types of 

neutron badges and two types of gamma badges.a

Szedziewski 1965

Late 1969 Film badge fading study. Author unknown 

ca. 1969

January 1973 Memorandum on dose rate for stress test facility. Holliday 1973

July 1974 Began using Landauer as source of film badges. Ward 1974

April 1978 Memorandum on personnel neutron dosimetry recommending use 

of new Landauer neutron badge using polycarbonate plastic.

Holliday 1978

October 1979 Began using Mound neutron dosimeters. Burkhart 1987b

1986 Landauer switched from reporting photon doses as exposures to 

reporting them as deep doses.

Yoder 2005

October 1986 Memorandum on estimated doses to GEND personnel handling 

unmarked neutron generator units.

Burkhart 1986

October 1987 End use of Mound neutron dosimeters. Burkhart 1987b

October 1988 Memorandum on radiation dose rates from RTG heat sources. Weaver 1988

April 1990 Changed from Landauer film to thermoluminescent dosimetry. Hall 1989

February 1991 All plutonium, with the exception of calorimeter sources and very 

small instrument calibration check sources, was removed from the 

Pinellas Plant by February 1991.

Author unknown 

undated c, MMSC 

1992

1971–1993 Various determinations for doses from testing of sealed neutron 

generators.

GE 1971–1996

October 1994 The primary Defense Program mission for the Pinellas Plant ended 

on October 1, 1994. The Pinellas Plant’s new mission was to clean 

up the facility from the past DOE mission and transition the site to 

an alternate use. 

MMSC 1994

a. The Pinellas Plant was the progeny of GEXM. Because the radiological control programs for these two sites were 

closely linked and because the available information indicates that the Pinellas Plant managed the radiological programs 

at GEXM after its startup, the information in this evaluation is considered to be applicable to the Pinellas Plant 

dosimeters.
 

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Figure 6-4. Redacted example of a Landauer dosimetry report from 1978. 
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Tables 6-4 through 6-9 summarize the monitoring techniques and describe the suspected and known 

Pinellas Plant dosimeters. 

Table 6-4. Whole body beta-gamma dosimeters used at Pinellas Plant.

a

Period Dosimeter type Dosimeter description

1957–Jun 1974 Photographic film

(beta-gamma) 

Photographic film dosimeters during this period were processed inhouse (Burkhart 1987a). Based on the early film worksheets that 

are available, the early beta-gamma dosimeters consisted of only 

an open window (OW) dosimeter and a single shielded dosimeter. 

Additionally, those records also indicate that only the OW dosimeter 

was normally read, unless a significant OW reading was recorded. 

By December 1970, the beta-gamma dosimeters appear to have 

consisted of an open window dosimeter and three shielded 

dosimeters. The shielded dosimeters appear to have utilized an 

aluminum (Al) filter and two cadmium (Cd) filters of different 

thicknesses. Based on the records for the area dosimeters, the 

filters were likely 0.040 in Al, 0.014 in Cd, and 0.040 in Cd. 

However, it still appears that only the OW dosimeter was normally 

read (GE 1957–1990, 1958–1973, and 1972).

Jul 1974–Mar 1990 Landauer Type Gb,c

(beta-gamma)

The Type G dosimeter was a film emulsion package placed in 

standard Gardray holder/badge for monitoring beta, X-ray, and 

gamma exposures. It was insensitive to neutron radiation. 

Required in areas where Kr-85 was used. Required for radiationgenerating equipment and accelerator operators.

Gamma and X-ray: 30 keV to 20 MeV; beta: over 1.5 MeV.

Apr 1990–1997 Landauer Type Z/Fc

(beta-gamma)

The Type Z dosimeter was comprised of 3 TLD-700 chips for 

monitoring beta, X-ray and gamma exposures. It was insensitive to 

neutron radiation. After 1994, the Type Z dosimeter was renamed 

the Type F dosimeter.

a. Sources: Landauer 1956–2002; Burkhart 1987b; GE 1986–1988, GE 1990, Burkhart 1990; Greene 1985a; Hall 1989; 

Holliday 1978; GE 1974–1980, Landauer 2005–2006; Ingle 1991; Weaver 1987, 1991, 1995 , 1996; ORAUT 2004a. 

b. A June 1974 memorandum indicates that the Pinellas Plant ordered Landauer Type J (beta-gamma) and Type K (betagamma and neutron) dosimeters (Burkhart 1987b); however, the actual dosimetry records indicate that Landauer Type 

G (beta-gamma) and Type P (beta-gamma and neutron) dosimeters were actually received and used (Landauer 1974–

1980). 

c. The use of the Landauer dosimeters is indicated by exposure type codes in the earlier years and use codes in the later 

years. The following are the dosimeter exposure type or use codes: 1 - whole body, 2 – skin/lens of eye, 3 - right finger, 

4 – left finger, 5 – right wrist, 6 – left wrist, 7 – other extremity, 8 – other whole body, 9 – monitor. 
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Table 6-5. Whole body neutron and beta-gamma-neutron dosimeters used at Pinellas Plant.

a

Period Dosimeter type Dosimeter description

1957– Jun 1974 NTA film

(neutron only)

Photographic film dosimeters to measure beta-gamma radiation 

were also equipped with NTA film to measure neutron radiation. As 

with the early beta-gamma dosimeters, these dosimeter were also

processed in-house (Burkhart 1987a). Fast neutrons undergoing 

elastic collision with content of emulsion or cellulose acetate base 

material produced recoil protons, which were recorded as 

photographic tracks in the NTA film emulsion. Track density was a 

linear function of dose. Developed images exhibited tracks caused 

by neutrons, which could be viewed and counted using appropriate 

imaging method (typically using a microscope or a magnified 

projector). Initially, total neutron tracks per 100 fields were counted. 

By January 1968, the total neutron tracks per 10 fields and 

sometimes neutron tracks per 50 fields were counted (GE 1957–

1990, 1958–1973, and 1972).

Jul 1974–May 

1978

Landauer Type Pb,c

(beta-gammaneutron)

The Landauer Type P dosimeter appears to have been used in all 

areas with a potential for neutron exposures (GE 1974–1980). The 

Type P dosimeter was a combination beta-gamma and fast neutron 

dosimeter. The fast neutron dosimeter was an NTA film dosimeter 

and the beta-gamma dosimeter was likely a Landauer Type G film 

dosimeter. 

Jun 1978–1997 Landauer Type Ec

(neutron only)

The Landauer Type E neutron dosimeter was used in all areas at 

the Pinellas Plant until October 1979 when the site started to use 

the Mound dosimeters in areas with RTG operations. After October 

1979, the Type E dosimeter was used only in the neutron generator 

areas at the Plant. 

The Type E dosimeter is a polycarbonate (Lexan) neutron recoil 

track registration device used to monitor fast neutron interactions. 

Neutrak 144 has a dosimeter element for response to fast neutrons. 

Neutrak E1 has a polyethylene radiator over CR-39 chip that would 

monitor for fast neutrons; only Lexan responded to neutrons by 

recording ionization damage caused by neutrons interacting with 

carbon and oxygen atoms, which leaves a track. It had uniform 

energy response from 3 to over 14 MeV with threshold of about 

1 MeV (Weaver 1987). 

The Type E dosimeter was combined with the Type G dosimeter 

(and later with the Type Z/F dosimeter) (Weaver 1987, 1996). 

Workers were required to wear an E/G dosimeter combination when 

working around neutron generators. Accelerator operators were 

also required to wear an E/G dosimeter combination. An E/G 

dosimeter combination or a G dosimeter was required when 

working with calibration sources (Weaver 1987). 

Oct 1979–Sep 

1987d

Mound TLD

(gamma-neutron)

The Mound TLD was used in areas with RTG operations (i.e., 

where PuO2 was handled). It was used to measure exposures to 

the X-rays and 2-MeV average neutrons from the handling of the 

sealed PuO2 sources during the production of RTG units in Building 

400. This dosimeter utilized a Harshaw 8810 TLD package that 

included an albedo neutron monitoring capability. The Harshaw 

8810 TLD package utilized a combination of TLD-600 and TLD-700 

dosimeter chips, which were encased in a plastic holder made of 

Cycolac.
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Period Dosimeter type Dosimeter description

Sep 1987d–Feb 

1991

Landauer Type R/Ic

(beta-gammaneutron)

The Landauer Type R dosimeter arrangement was used in areas 

with RTG operations (i.e., where PuO2 was handled), and replaced 

the Mound TLD. The Type R dosimeter package consisted of a 

Type G dosimeter for measuring beta, X-ray, and gamma radiation 

and a Type I dosimeter for measuring neutron radiation. The Type I 

dosimeter combined a TLD albedo neutron monitor with a track 

recoil device (CR-39 [allyl diglycol carbonate]) that responds to 

proton recoil events. Neutron energy range was approximately 

1 x 10-6 to 10 MeV. Albedo response to thermal neutron radiation 

was subtracted to yield fast neutron dose. The “Neutrak ER” has an 

albedo element with the above-described elements. A qualitative 

relationship was derived to determine ratios of neutrons of various 

energies. When first used at the Pinellas Plant, the Type I 

dosimeters did not meet all DOELAP requirements during 

performance testing (Weaver 1991). 

After the Type G dosimeter was replaced by the Type Z dosimeter, 

the combined dosimeter unit was known as the Type I dosimeter. 

After production of RTGs was halted in October 1991, the Type R/I 

dosimeter was used only as an area monitor for the americiumberyllium (AmBe) source (Weaver 1991).

a. Sources: Landauer 1956–2002; Burkhart 1987b; GE 1986–1988, GE 1990, Burkhart 1990; Greene 1985a; Hall 1989; 

Holliday 1978; GE 1974–1980, Landauer 2005–2006; Ingle 1991; Weaver 1987, 1991, 1995 , 1996; ORAUT 2004a. 

b. A June 1974 memorandum indicates that the Pinellas Plant ordered Landauer Type J (beta-gamma) and Type K (betagamma and neutron) dosimeters (Burkhart 1987b); however, the actual dosimetry records indicate that Landauer Type 

G (beta-gamma) and Type P (beta-gamma and neutron) dosimeters were actually received and used (Landauer 1974–

1980). 

c. The use of the Landauer dosimeters is indicated by exposure type codes in the earlier years and use codes in the later 

years. The following are the dosimeter exposure type or use codes: 1 - whole body, 2 – skin/lens of eye, 3 - right finger, 

4 – left finger, 5 – right wrist, 6 – left wrist, 7 – other extremity, 8 – other whole body, 9 - monitor.

d. Dual Mound and Landauer Type R neutron dosimeters were used to monitor RTG workers during the month of 

September 1987 (Burkhart 1987b).

Table 6-6. Wrist beta-gamma dosimeters used at Pinellas Planta. 

Period Dosimeter type Dosimeter description

Jul 1974–1990 Landauer Type Gb,c

(beta-gamma)

The Type G wrist dosimeter was a film emulsion package placed in 

a standard Gardray holder/badge for monitoring beta, X-ray, and 

gamma exposures. It was insensitive to neutron radiation. The 

Type G wrist dosimeter was used to monitor extremity doses in 

areas with RTG operations (i.e., where PuO2 was handled) (Weaver 

1987); extremity dosimetry was assigned in other areas as needed 

(Weaver 1996).

1991–1997 Landauer Type Kb

(beta-gamma)

The Landauer Type K wrist dosimeters assigned during this period 

utilized three TLD-100 chips.

a. Sources: Landauer 1956–2002; Burkhart 1987b; GE 1986–1988, GE 1990, Burkhart 1990; Greene 1985a; Hall 1989; 

Holliday 1978; GE 1974–1980, Landauer 2005–2006; Ingle 1991; Weaver 1987, 1991, 1995 , 1996; ORAUT 2004a. 

b. The use of the Landauer dosimeters is indicated by exposure type codes in the earlier years and use codes in the later 

years. The following are the dosimeter exposure type or use codes: 1 - whole body, 2 – skin/lens of eye, 3 - right finger, 

4 – left finger, 5 – right wrist, 6 – left wrist, 7 – other extremity, 8 – other whole body, 9 - monitor.

c. A June 1974 memorandum indicates that the Pinellas Plant ordered Landauer Type M (beta-gamma) wrist dosimeters 

(Burkhart 1987b); however, the actual dosimetry records indicate that Landauer Type G (beta-gamma) wrist dosimeters 

were actually received and used (GE 1974–1980).
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Table 6-7. Ring beta-gamma dosimeters used at Pinellas Planta. 

Period Dosimeter type Dosimeter description

1957–1974 Photographic film

(beta-gamma) 

Outside of the fact that the photographic film dosimeters during this 

period were processed in-house (Burkhart 1987a), no other details 

about these dosimeters are known.

1975–1983 (unknown) This analysis found no information regarding the type of ring 

dosimeters used during this period. If ring dosimeters were used 

during this period, they might have included the in-house film 

dosimeters or Landauer Type L ring dosimeters. Because the type 

of ring dosimeter that might have been used is uncertain, extremity 

doses based on ring dosimeters from this period will need to be 

assessed on a case-by-case basis.

About 1983b–1997 Landauer Type Uc

(beta-gamma)

The Landauer Type U ring dosimeters assigned during this period 

utilized one LiF TLD chip for monitoring beta (>1.5 MeV), X-ray, and 

gamma exposures. It was insensitive to neutron radiation. The 

Type U ring dosimeter was used to monitor extremity doses in 

areas with RTG operations (i.e., where PuO2 was handled) (Weaver 

1987); extremity dosimetry was assigned in other areas as needed 

(Weaver 1996).

a. Sources: Landauer 1956–2002; Burkhart 1987b; GE 1986–1988, GE 1990, Burkhart 1990; Greene 1985a; Hall 1989; 

Holliday 1978; GE 1974–1980, Landauer 2005–2006; Ingle 1991; Weaver 1987, 1991, 1995 , 1996; ORAUT 2004a. 

b. This analysis found no documentation that shows the start of use for the Landauer Type U finger ring dosimeters. 

c. The use of the Landauer dosimeters is indicated by exposure type codes in the earlier years and use codes in the later 

years. The following are the dosimeter exposure type or use codes: 1 - whole body, 2 – skin/lens of eye, 3 - right finger, 

4 – left finger, 5 – right wrist, 6 – left wrist, 7 – other extremity, 8 – other whole body, 9 - monitor.

6.2.4 Dosimetry Performance and Calibration

6.2.4.1 Performance Testing

During 1965, the GE X-Ray Division in Milwaukee (GEXM) compared the performance of the gamma 

and neutron dosimeters being used at GEXM to dosimeters with commercial services from R. S. 

Landauer Jr. & Company (Landauer), which showed essentially equivalent performance (Szedziewski 

1965). Because the gamma and neutron dosimeters being used at GEXM were being processed by 

the Pinellas Plant (Burkhart 1987a), this performance comparison is also applicable to the Pinellas 

Plant’s dosimeter performance.

As indicated in Table 6-5, NTA film was used for the neutron dosimeters from 1957–May 1978 (Author 

unknown ca. 1969; ORAUT 2004a). As stated in the Hanford Site, Idaho National Laboratory, and 

Nevada Test Site Occupational External Dose TBDs (ORAUT 2010, 2011a, and 2012a), NTA film was 

basically the only common dosimeter method available to measure neutron dose in AEC facilities prior 

to the use of TLDs. The neutron spectra at Pinellas were known to be dominated by higher energy 

14-MeV deuterium-tritium and 2.5 MeV deuterium-deuterium fusion neutrons due to the unique 

designs of the neutron generators. These higher-energy neutron fields tend to minimize errors from 

fading or energy response reported at the first AEC Neutron Dosimetry Workshop in 1969 which 

indicated that Savannah River Site calibration laboratory dose measurements made with NTA film 

were about one-half to one-fourth of those measured with other methods, including the neutron TLD 

(Vallario, Hankins, and Unruh 1969).

A circa 1969 study, indicated that the NTA film neutron dosimeters experienced track fading during 

use, which caused a loss of information (Author unknown ca.1969, Holliday 1978). The maximum 

errors associated with the use of this factor occurred when a worker received a majority of the neutron 

exposure at the beginning or the end of a dosimeter monitoring period. The assigned dose to a 

worker receiving a total exposure on the first day of a dosimeter period would be 20% of the true dose 

per week; while the dose assigned to a worker receiving a total exposure on the last day of the 
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monitoring period would be 40% of the true dose for a monthly wear period (Author unknown ca. 

1969). As a result of that study, GEND Health Physics started applying a factor of 3 to correct for the 

track fading in January 1970 (Author unknown ca. 1969). The Landauer polycarbonate badge that 

replaced these badges in May 1978 were not subject to track fading, so no correction factor was 

applied after May 1978 (Holliday 1978). 

The GEND Health Physics Group determined that the Neutrak ER portions of the Landauer Type R/I 

neutron dosimeter recorded only 67% of total neutron dose because of the dosimeter’s poor energy 

response to low-energy neutrons from the RTG PuO2 heat sources (Burkhart 1987a).

6.2.4.2 Calibration

Pinellas conducted dosimetry calibration as part of its external dosimetry audit program. This type of 

performance testing occurred every 6 months using radiation sources with known strengths. Tests 

used the DOE Standard for the Performance Testing of Personnel Dosimetry Systems (DOE 1986) 

and American National Standards Institute ANSI/HPS13N.11, Personal Dosimetry Performance – 

Criteria for Performance (HPS 1993). Each test used approximately six to nine badges.

The photon calibration of the Landauer Type G dosimeters was performed by exposing the 

dosimeters to the Shepherd Model 81-12 137Cs beam irradiator in Building 800 (a gamma check for 

662-keV photons). For beta calibration, the dosimeters were exposed by placing them on a bare 

uranium slab for exposure to the resulting radiation. A covering with a known density thickness was 

placed on the dosimeters to keep them free from uranium contamination (GE 1990).

Calibration of the Landauer Type E polycarbonate dosimeters was performed by exposing them to a 

D-T fast neutron source with known source strength. The badges were placed on a Lexan “jig” and 

set at a known distance from the source of neutrons (GE 1990).

The Landauer Type R dosimeter, which consisted of a beta-gamma film dosimeter, an albedo neutron 

monitoring TLD, and a CR-39 neutron dosimeter, was placed on a water phantom and exposed to a 

Shepherd Model 149 calibrator equipped with an 241AmBe source. The phantom was level with (and 

at known distances from) the source on a moveable metal rack about 4 ft above the floor to minimize

scattering effects (GE 1990).

However, even though there is Pinellas Plant documentation showing such calibration studies 

occurred, the results of the studies and subsequent use in the radiation dosimetry program are not 

available.

Beginning in 1974, Landauer supplied all dosimetry badges and performed the necessary calibrations. 

Landauer used control film. The personnel monitoring reporting was normally in net exposure; the 

control film reading was deducted from the personnel film reading. If the control film appeared to 

have been exposed differently than the personnel packets, the densities on the personnel film were 

normalized to Landauer controls only and a nonminimal control reading was reported. A control 

packet reading was provided in arbitrary units, not necessarily in millirem. Minimal beta or soft X-ray 

skin dose readings were not reported until after a positive skin dose exposure was recorded. Ring 

badges were calibrated only for high-energy gamma (probably >0.662 MeV) and high-energy beta 

(1.5 MeV) unless special arrangements were made with the Plant (ORAUT 2004a). As indicated in 

Table 6-3, Landauer switched from reporting photon doses as exposures to reporting them as deep 

doses in 1986 (Yoder 2005).

Further details on the dosimetry used at the Pinellas Plant are provided in Sections 6.3 and 6.4. 
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6.2.5 Historical Administrative and Reporting Practices

6.2.5.1 Administrative Practices

The Pinellas Plant started an external dosimetry program in 1957 to monitor individual employees 

working in neutron generator production areas. Some Pinellas records on facility monitoring, safety 

evaluations, investigations, etc., exist; however, most of these records concern operations after 1970. 

Records of radiation dose to individual workers from personnel dosimeters are generally available for 

1957 to 1994 for the workers’ time of employment. The dose measured by dosimeters was recorded 

at the time of measurement, reviewed by Pinellas Plant health physicists, and routinely made 

available to workers. The guidance document External Dose Reconstruction Implementation 

Guidelines, OCAS-IG-0001 (NIOSH 2007), indicates that these records represent the highest quality 

record for a retrospective dose assessment. The information in this section pertains to the analysis of 

the available records.

Table 6-8 lists the total number of Pinellas Plant employees and the numbers of employees who were 

monitored for radiation exposure for years that this type of data was available. From 1960 to 1973, 

U.S. Atomic Energy Commission (AEC) annual exposure summary reports indicate that Pinellas had 

27.5% of its labor force wearing external dosimetry. For the years through 1989, external dosimetry 

was exchanged and analyzed monthly. Beginning in January 1990, external dosimetry was 

exchanged and analyzed quarterly (Burkhart 1988; Weaver 1990).

As indicated in Table 6-8, not all Pinellas Plant employees were monitored for external dose. This 

was due to the fact that the Pinellas Plant was not required to monitor most employees, because the 

majority of the work performed at the Pinellas Plant did not involve exposures to external sources of 

radiation and because most personnel did not have routine access to radiation areas. 

The following are some examples of the historical guidelines and regulations for radiation protection 

and worker dose monitoring requirements that were effective throughout the Pinellas Plant’s 

operational history. One of the earliest guidelines on radiation protection was the 1958 Addendum to 

National Bureau of Standards Handbook 59, which recommended that it be made improbable for any 

individual outside of a controlled area to receive a dose of more than 0.5 rem/year from external 

radiation (NBS 1958). The first radiation protection regulation was the AEC’s initial issue of 10 CFR 

20 in 1960, which imposed external dose limits of 1.25 rem/quarter for restricted areas and no more 

than 0.100 rem in any seven consecutive days for unrestricted areas (AEC 1960). In 1974 the AEC 

was abolished and its functions were divided into the newly formed ERDA and NRC. After that 

transition, 10 CFR 20 became a regulation that was only applicable to NRC facilities. After ERDA was 

reorganized into the DOE in 1977, the requirements for DOE facilities were driven by DOE directives 

and orders. The 1981 version of DOE Order 5480.1 required workers to be monitored for external 

dose if they had the potential to receive more than 10% of the quarterly dose limit (i.e. more than 

0.300 rem/quarter) (DOE 1981). Since 1988, DOE Order 5480.11 (DOE 1988) and later 10 CFR 835 

have required workers with a potential to receive 0.100 rem/year of external dose to be monitored.
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Table 6-8. Annual whole body radiation exposure information for the Pinellas 

Plant .a,b

Year

Number of

Pinellas Plant

employees

Number of 

monitored 

employees

Number of 

annual doses 

<1 rem

Number of 

annual doses 

1–2 rem

1960 1,304 225 225 0

1961 1,395 251 251 0

1962 1,370 254 254 0

1963 1,597 545 545 0

1964 1,408 347 347 0

1965 1,319 301 301 0

1966 1,445 325 325 0

1967 1,405 585 584 1

1968 1,424 292 292 0

1969 1,323 588 588 0

1970 1,311 442 441 1

1971 1,283 410 410 0

1972 1,402 346 346 0

1973 1,252 383 383 0

1974 Not available Not available Not available Not available

1975 Not available Not available Not available Not available

1976c Not available 317 317 0

1977 Not available 300 300 0

1978 Not available 298 298 0

1979 Not available 334 334 0

1980 Not available 376 376 0

1981 Not available 389 389 0

1982 Not available 408 408 0

1983 Not available 378 378 0

1984 Not available 405 405 0

1985 Not available 391 391 0

a. Source: ORAUT 2017a. 

b. The values in this table include all AEC/ERDA/DOE and GE employees. They do not 

include any data for Pinellas Plant visitors.

c. The values for 1976–1985 include the numbers of workers monitored for external dose 

and/or internal dose, and their reported annual doses. Whereas, the values for 1960–1973 

are limited to only external dose information. Additionally, the reported information for 

1976–1985 no longer included the numbers of unmonitored employees, so the total 

number of employees could not be determined.

General operating procedures for the Pinellas Plant from as early as 1967 indicate that dosimeters 

were assigned to all personnel with a potential to receive a measurable external dose. Pinellas Plant

general operating procedure titled “Film Badge Requirements,” dated May 10, 1967, states that it was 

management’s responsibility to “Ascertain that film badges are worn by all personnel whose work 

assignments entail the potential for accumulating measurable radiation exposures; compliance with 

this procedure is a condition of employment in a Radiation Area.” (GE 1957–1990 p. 127–128). All 

versions of a Pinellas Plant general operating procedure titled “Assignment of Personnel to Work in 

Radioactive Material, Contamination, or Radiation Areas,” dated between January 17, 1968, and June 

29, 1984, state that it was the Work Area Manager’s responsibility to “Ensure that all employees 

having work assignments in Radiation Areas are provided with film badges.” (GE 1957–1990 p. 122–

125 and 129–130). In addition, a Pinellas Plant memorandum dated November 1, 1984, states that 

“All individuals with a potential of receiving measurable radiation dosage are included in our personnel 

monitoring program.” (Greene 1984a). Based on a comparison of these procedures to the historical 

guidelines and regulations, the Pinellas Plant monitored significantly more workers for external dose 

than what was necessary. 
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The reported annual whole-body doses indicate that on average approximately 95% of the monitored 

workers at the Plant received an annual whole-body dose of less than or equal to 0.100 rem for any 

given year (see Table B-1 in Attachment B). Table B-1 also indicates that nearly 80% of those same 

workers actually received annual doses ≤ 0.020 rem on average. This confirms that the Pinellas Plant 

did monitor any personnel with the potential to receive a measurable external dose, which exceeded 

the worker monitoring requirements throughout its entire history. 

A Pinellas Plant health physicist reviewed the reports and evaluated and resolved unusual or 

inconsistent results. The health physicist could modify the reports, documenting all investigations and 

reasons for such modifications. These reports were placed in the worker’s dosimetry file. The health 

physicist checked the printed version of the Landauer reports against the electronic version. Until 

1990, workers who reported lost or stolen badges were assigned an exposure that was an average 

from their previous dose histories (GE 1990).

If Landauer found that a badge exceeded 400 mrem whole body, 800 mrem skin, or 6,000 mrem 

extremity, it was required to call the responsible Pinellas health physicist (GE 1990). Analysis of the 

available worker records found no documentation that this reporting requirement was ever exceeded 

during the time Landauer provided dosimetry services to the Plant (1974 to 1997).

6.2.5.2 Reporting Practices 

Based on the available claim records, the Pinellas Plant routinely recorded cumulative career dose 

totals, annual dose totals, and individual dosimeter results for the monitored workers. Attachment A 

contains examples of some of the external dosimetry records are provided by the DOE for the 

EEOICPA claims. The examples in Attachment A only include examples of the records that provided 

individual dosimeter results for the workers. It should be noted that the reported monthly, annual, 

cumulative, and lifetime dose totals for the Pinellas Plant were reported as whole body doses, which 

would include any internal tritium doses that the worker received. 

By the late-1980’s, the Pinellas Plant had begun an “As Low As Reasonably Achievable” (ALARA) 

Program, which provided another layer of tracking and reporting worker doses (Burkhart 1989, 

Weaver 1994b, LMSC 1996). 

6.3 ADJUSTMENTS TO RECORDED DOSE

The following adjustments need to be made to the reported doses from certain dosimeters used at the 

Pinellas Plant, to account for biases that may have resulted in under-measurements of the workers’ 

external doses.

The recorded Pinellas doses show that there is not a consistent relationship between recorded 

neutron and photon doses for work performed in the neutron generator areas at the Plant. This lack 

of a true neutron-to-photon dose ratio can be attributed to the nature of the Pinellas processes, during 

which neutron generator testing occurred in open rooms that, combined with the short period of the 

neutron pulse, relatively open test structure, virtually no photons from the neutron generator, and 

relatively low number of neutrons per test pulse (no significant quantities of activation products), 

would result in a corresponding photon dose. This result is supported in the individual dose records, 

which indicate that the timing of Pinellas personnel neutron and photon exposures varied greatly not 

only on a yearly but also on a monthly basis (a recorded value for 1 month and no recorded doses for 

the next several months). Thus, the assignment of a neutron-to-photon dose ratio to adjust for a 

missed neutron dose is not valid for the Pinellas Plant for neutron generators.

However, for RTG PuO2 heat sources an approximate 3:1 neutron-to-photon ratio was measured 

based on neutron and photon exposure rates from the processing of the RTG PuO2 heat sources as 
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measured in 1987 (GE 1987). Exposure rates measured at another time were 0.37 mR/hr neutron 

and 0.18 mR/hr gamma at 24 in. Actual dose equivalent received from removal and processing of 15 

units for one worker (apparently the typical workload was 50 generators per month for perhaps three 

personnel) amounted to a 20.9-mrem neutron and 7.1-mrem gamma dose equivalents or 

approximately a 3:1 neutron-to-photon ratio (GE 1983). Because the RTG radioisotopic ratios and 

activities remained unchanged and the workstations and processes consistent, this ratio can also be 

considered a constant for RTG work conducted from 1975 to 1990.

6.3.1 Photon Dose Adjustments

With the exception of the Mound TLDs, no adjustments to photon doses are necessary for dose 

reconstruction.

6.3.1.1 Signal Fading Adjustments

Prior to July 1981, the Mound TLDs were subject to signal fading, which could amount to 25 to 30% of 

the stored signal fading during a period of 4 weeks (Crain 1981a). Beginning with the third quarter of 

1981, the Mound TLDs were annealed before being packed into the dosimeter holders/badges to 

reduce the signal fading problem to a negligible level (Crain 1981a, 1981b). As a result of this signal 

fading, the reported doses for the affected dosimeters were as little as 70% of the actual dose. To 

compensate for the signal fading, the Mound TLD results reported during the period from October 

1979 through June 1981 should be multiplied by a signal fading correction factor of 1.43 (1/0.70). 

After June 1981, there was no significant signal fading for those dosimeters, and no corrections for 

signal fading are necessary.

6.3.2 Neutron Dose Adjustments

The following subsections provide the bases for the various neutron dose adjustments. The 

necessary neutron dose adjustments are summarized in Tables 6-9, 6-10, and 6-11 below.

6.3.2.1 Radiation Weighting Factor Adjustments

The Pinellas Plant used a relative biological effectiveness (RBE) weighting factor of 10.0 when 

calculating the effective dose for 14-MeV neutrons (Holliday undated), which is equivalent to the ICRP 

Publication 60 neutron weighting factor for neutron energies from 2 to 14 MeV (ICRP 1991). Because 

the weighting factor value used by the Pinellas Plant is higher than other values that could have been 

used (such as those in NCRP 1971), the ICRP Publication 60 correction factor is unity for neutron 

exposures in the neutron generator areas.

The plutonium in the RTGs emits neutrons with an average energy of 2 MeV and an energy range of 

thermal to 12 MeV (Figure 6-3) (Burkhart 1987a). Based on the average neutron energy being 

2 MeV, 50% of the neutrons were assumed to be in the 2- to 20-MeV neutron energy group and the 

remaining 50% of the neutrons were assumed to be in the neutron energy group of 0.1- to 2-MeV, 

which is typically the most favorable-to-claimant neutron energy group. The available GEND 

documentation does not describe any separate processing of the dosimetry when Landauer 

dosimeters were used to monitor exposures from RTG operations. Therefore, it was assumed that an 

RBE of 10 was applied to the neutron doses from RTG operations when Landauer dosimeters were 

used for the RTG operations (i.e., 1975–September 1979 and September 1987–February 1991). 

When the Mound dosimeters were used to monitor exposures from RTG operations (i.e., October 

1979–September 1987), a single RBE value of 7 was applied to the Pinellas Plant doses reported by 

the Mound Laboratory (ORAUT 2004). Because the RBE value used for the Landauer dosimeters is 

uncertain and because the RBE used for the Mound dosimeters is likely more favorable to claimants 
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than what was used for the Landauer dosimeters, the radiation weighting factor adjustments for all 

neutron energy groups and all periods of RTG operations should be based on an RBE value of 7.

A summary of the ICRP Publication 60 correction factors (ICRP 1991) for the Pinellas Plant’s reported 

neutron doses is provided in Table 6-9. 

Table 6-9. Neutron radiation energies and ICRP 60 correction factors. 

Process type

Neutron

energy range

IREP energy

group

Default dose

fractiona

(%)

ICRP 60/RBE 

ratiob

ICRP 60 

correction 

factorc

Neutron 

generator 

operations

≤14 MeV 2–20 MeV 100% 10/10 1.00

RTG 

operations

0–2 MeV 0.1–2 MeV 50% 20/7 1.43

RTG 

operations

2–12 MeV 2–20 MeV 50% 10/7 0.71

a. Because no neutron dose values for the various neutron energies were available at the time of this assessment, the 

default dose fractions for the Pinellas Plant are actually just based on the expected neutron energy distributions.

b. The ICRP 60/RBE ratio is the appropriate radiation weighting factor in ICRP 60 (ICRP 1991) divided by the historical 

RBE value that was applied to the reported neutron doses.

c. The ICRP 60 correction factor is the ICRP 60/RBE ratio multiplied by the appropriate fraction/percentage for the neutron 

energy group.

6.3.2.2 Track Fading Adjustments

The NTA track film that was used at the Pinellas Plant from the start of operations in 1957 through 

June 1978 was susceptible to track fading. Track fading occurs between the time the tracks in the 

dosimeter were created and when the dosimeter film was analyzed. As this time interval increases, 

the amount of track fading increases.

A study was performed in 1969 to determine the amount of track fading on the NTA track film 

dosimeters that were used at the Pinellas Plant. Track fading was potentially significant because of 

the Plant’s monthly dosimeter exchange frequency. The study determined that an average of 67% of 

the proton-recoil tracks had faded for a monthly dosimeter exchange. In other words, on average only 

33% of the original tracks remained by the time the dosimeters were analyzed. The study 

recommended that a correction factor of 3 (i.e., 1/0.33) be incorporated into the dose calculations 

beginning in January 1970 to account for track fading (Author unknown ca. 1969). A 1974 

memorandum from Landauer indicates that the correction factor of 3 was also being applied to the 

NTA film dosimeter results provided by Landauer for the period of July 1974 through May 1978 

(Wheeler 1974).

Because there is no indication that the Pinellas Plant performed any track fading corrections to the 

reported neutron doses prior to 1970, the dose reconstructor should apply a track fading correction 

factor of 3.0 to the neutron doses reported for the years of 1957–1969. No track fading corrections 

need to be applied to the neutron doses reported after 1969, because the reported neutron doses 

have already been corrected.

6.3.2.3 Signal Fading Adjustments

Prior to July 1981, the Mound TLD was subject to signal fading, which could amount to 25% to 30% of 

the stored signal fading during a period of 4 weeks (Crain 1981a). Beginning with the third quarter of 

1981, the Mound TLDs were annealed before being packed into the dosimeter holders/badges to 

reduce the signal fading problem to a negligible level (Crain 1981a, 1981b). As a result of this signal 

fading, the doses reported for the affected dosimeters were as little as 70% of the actual dose. To 
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compensate for the signal fading, the Mound TLD results reported during the period from October 

1979 through June 1981 should be multiplied by a signal fading correction factor of 1.43 (1/0.70). 

After June 1981, there was no significant signal fading for those dosimeters, and no corrections for 

signal fading are necessary.

6.3.2.4 Neutron Energy Response Adjustments

Because of the higher neutron energies associated with the neutron generator production activities, 

the reported neutron doses for workers in the neutron generator portions of the Pinellas Plant do not 

need to be adjusted for the dosimeters’ poor response to lower neutron energies. However, reported 

neutron doses for the Plant’s RTG workers need to be adjusted because of the limitations of some of 

the dosimeter responses to the lower energy neutrons that were encountered in the RTG areas.

NTA film has a neutron energy response threshold of 0.5 MeV. Because some of the neutrons 

associated with the RTG PuO2 heat sources had neutron energies below 0.5 MeV, the reported 

neutron doses from the start of Pinellas operation through June 1978 require a correction for poor 

energy response. Based on the neutron energy distribution information for the RTG PuO2 heat 

sources (Burkhart 1987a), 6.4% of the neutrons were in the 0–1 MeV energy range and were 

assumed to not have been detected by the NTA film dosimeters. Therefore, a neutron energy 

response correction factor of 1.07 [1/(1-0.064)] should be applied to all NTA film dosimeter results to 

account for any under-reported neutron doses due to the Nevada Test Site film dosimeter’s energy 

response.

Because the Mound dosimeters utilized TLD chips that responded well to the neutron energies 

encountered by RTG workers, no adjustment for the neutron energy response of the Mound 

dosimeters is necessary (ORAUT 2004b).

The GEND Health Physics Group determined that the Neutrak ER portions of the Landauer Type R/I 

neutron dosimeter recorded only 67% of total neutron dose because of the dosimeter’s poor energy 

response to low-energy neutrons from the RTG PuO2 heat sources (Burkhart 1987a). In addition, 

there is no indication that the reported neutron doses for the RTG workers have been adjusted by the 

Pinellas Plant. Therefore, dose reconstructors should use a neutron energy response correction 

factor of 1.49 (1/(0.67) to account for the underreported neutron doses from the RTG PuO2 heat 

sources.

Table 6-10. Summary of neutron dosimeter correction factors (CF) for 

neutron generator operations (1957–1997).

a,b

Applicable period

Track

Fading

CF

Signal

Fading

CF

Energy

Response

CF

Total

adjustment

1957–1969 3.00 N/A N/A 3.00

1970–1997 N/A N/A N/A 1.00

a. N/A = not applicable. 

b. This summary includes all of the neutron dosimeter correction factors except the 

ICRP Publication 60 correction factors, which are already summarized in 

Table 6-9 
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Table 6-11. Summary of neutron dosimeter correction factors (CF) for 

RTG operations (1975–1991).

a,b

Applicable period

Track

Fading

CF

Signal

Fading

CF

Energy

Response

CF

Total

adjustment

1975–Sep 1979 N/A N/A 1.07 1.07

Oct 1979–Jun 1981 N/A 1.43 N/A 1.43

Jul 1981–Sep 1987 N/A N/A N/A 1.00

Oct 1987–Feb 1991 N/A N/A 1.49 1.49

a. N/A = not applicable

b. This summary includes all of the neutron dosimeter correction factors except the 

ICRP Publication 60 correction factors, which are already summarized in

Table 6-9. 

6.3.3 Electron Dose Adjustments

Electron doses were monitored but not routinely recorded. The primary source of electron exposures 

was from the use of two85Kr leak detection systems (Radiflo and TRACER-flo systems) in Area 109 

from about 1963 through 1994.

Beta dose monitoring for 85Kr started before the DOELAP standard release, which used a calibration 

factor from a 90Sr/90Y source that tended to underestimate the dose from 85Kr exposures. To 

compensate for the lower energy of 85Kr in relation to that of 90Sr/90Y, a correction factor might have 

been used for the Pinellas Plant based on the more similar 204Tl energy spectrum. Because it is not 

clear from the Plant records whether the 204Tl energy calibration or equivalent was requested by 

Pinellas of Landauer or other vendors prior to 1986, a correction should be performed on the reported 

non-penetrating or electron doses prior to 1986 to ensure that the workers’ doses are not 

underestimated. To compensate for the energy spectrum differences the reported nonpenetrating or 

electron doses should be multiplied by a correction factor of 3.5 for the years prior to 1986 (Poliziani 

1985). From 1986 onwards, when DOELAP and National Voluntary Laboratory Accreditation 

Program standards included 204Tl calibration criteria for 85Kr exposures, it is not necessary to apply a 

correction factor.

6.4 MISSED DOSE

Missed external dose is the unrecorded or unmeasured external dose that is the result of either 

relatively high detection limits, short monitoring periods, high dosimeter exchange frequencies, or a 

combination of these three factors (NIOSH 2007). Missed doses are applicable to the Pinellas Plant 

workers that were monitored for external dose and had one or more reported dosimeter readings that 

were less than half the LOD for the dosimeter. Missed dose is primarily estimated based on 

dosimeter results, where the number of zero or < LOD/2 values for a given year is multiplied by the 

LOD/2 value for the dosimeters used during that year (NIOSH 2007).

At the Pinellas Plant, electron, photon, and neutron doses were possible. However, electron doses, 

including missed electron doses, were unlikely, as discussed in Section 6.2.2.1. Based on the 

available information for the Pinellas Plant, chronic external exposures to electrons did not occur and 

acute external exposures to electrons were unlikely. Therefore, missed electron doses do not 

normally need to be assessed for Pinellas Plant workers.

Tables 6-12 through 6-17 summarize the dosimetry parameters to be used for calculating missed 

doses for Pinellas Plant workers. For the calendar years that have more than one dosimeter LOD 

value, the highest LOD value should be used for the entire year, unless all of the dosimeter results are 

for the period with the lower LOD value.
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Table 6-12. Whole-body beta-gamma dosimetry missed doses for Neutron Generator Operations 

(1957–1997).a

Dosimeter type Period of use LOD (rem)

Exchange 

frequency

Maximum annual missed

dose (rem)b

Photographic film Jan 1957–Jun 1974 0.040 photonsc

0.040 electronsc 

Monthly 0.240 photons

0.240 electrons

Landauer Type G/Pd Jul 1974–Apr 1990 0.010 photonse

0.040 electronse

Monthly

(1974–1989)

0.060 photons

0.240 electrons

Landauer Type G/Pd Jul 1974–Apr 1990 0.010 photonse

0.040 electronse

Quarterly

(after 1989)

0.020 photons

0.080 electrons

Landauer Type Z/Ff May 1990–Dec1997 0.010 photonse

0.040 electronse

Quarterly 0.020 photons

0.080 electrons

a. Some of the beta-gamma dosimeters listed in this table included a component for monitoring neutron exposures; 

however, only the beta-gamma components of those dosimeters are addressed in this table.

b. The maximum annual missed dose assignments are calculated by multiplying half of the dosimeter’s LOD value by the 

maximum number of dosimeters exchanged in a given year (NIOSH 2007).

c. Estimated LODs for commonly used photon dosimetry (ORAUT 2006b). 

d. The Landauer Type P dosimeter was a beta-gamma-neutron dosimeter that likely utilized a Type G dosimeter or 

equivalent for its beta-gamma component.

e. The photon and electron LODs for this Landauer dosimeter are based on information in GE (1974-1980).

f. After 1994, the Landauer Type Z dosimeter was renamed the Type F dosimeter. 

Table 6-13. Whole-body beta-gamma dosimetry missed doses for RTG Operations (1975–1991).a

Dosimeter type Period of use LOD (rem)

Exchange 

frequency

Maximum annual missed

dose (rem)b

Landauer Type G/Pc Late-1975–Sep 1979 0.010 photonsd

0.040 electronsd

Monthly 0.060 photons

0.240 electrons

Mound TLDe Oct 1979–Jun 1981 0.029 photonsf,g Monthly 0.174 photons

Mound TLDe Jul 1981–Aug 1987h 0.020 photonsf Monthly 0.120 photons

Landauer Type R/I Oct 1987h–Feb 1991 0.010 photonsd

0.040 electronsd

Monthly

(1987–1989)

0.060 photons

0.240 electrons

Landauer Type R/I Oct 1987h–Feb 1991 0.010 photonsd

0.040 electronsd

Quarterly

(after 1989)

0.020 photons

0.080 electrons

a. Some of the beta-gamma dosimeters listed in this table included a component for monitoring neutron exposures; 

however, only the beta-gamma components of those dosimeters are addressed in this table.

b. The maximum annual missed dose assignments are calculated by multiplying half of the dosimeter’s LOD value by the 

maximum number of dosimeters exchanged in a given year (NIOSH 2007).

c. The Landauer Type P dosimeter was a beta-gamma-neutron dosimeter that likely utilized a Type G dosimeter or 

equivalent for its beta-gamma component.

d. The photon and electron LODs for this Landauer dosimeter are based on information in GE (1974-1980).

e. The Mound TLD was a gamma-neutron dosimeter.

f. The LOD for this Mound neutron dosimeter was 0.020 mrem (ORAUT 2004).

g. The LOD for this dosimeter has been adjusted to account for signal fading, in accordance with Section 6.3.1.1, and that 

adjustment is shown in ORAUT (2017b). 

h. Dual Mound and Landauer Type R neutron dosimeters were actually used to monitor RTG workers during the month of 

September in 1987 (Burkhart 1987b). Because it is uncertain which dosimeters were used for the reported doses, the 

dosimeter with the most favorable-to-claimant LOD values was used and the period of use date has been adjusted to 

reflect that. 
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Table 6-14. Whole-body neutron dosimetry missed doses for Neutron Generator Operations (1957–

1997).

a

Dosimeter

type Period of use LOD (rem)

Exchange

frequency

Maximum annual

missed dose (rem)b

NTA film Jan 1957–Dec 1969 0.150 neutronsc,d Monthly 0.900 neutrons

NTA film Jan 1970–Jun 1974 0.050 neutronsc Monthly 0.300 neutrons

Landauer Type Pe Jul 1974–May 1978 0.020 neutronsf Monthly 0.120 neutrons

Landauer Type E Jun 1978–Dec 1997 0.020 neutronsf Monthly 

(1974–1989)

0.120 neutrons

Landauer Type E Jun 1978–Dec 1997 0.020 neutronsf Quarterly 

(after 1989)

0.040 neutrons

a. Some of the neutron dosimeters listed in this table included a component for monitoring beta-gamma exposures; however, 

only the neutron components of those dosimeters are addressed in this table. 

b. The maximum annual missed dose assignments are calculated by multiplying half of the dosimeter’s LOD value by the 

maximum number of dosimeters exchanged in a given year (NIOSH 2007).

c. NTA film dosimeters have an LOD of 0.050 rem for fast neutrons based on Wilson et al. (1990).

d. The LOD for this NTA film dosimeter has been adjusted to account for track fading, in accordance with Section 6.3.2.2, and 

that adjustment is shown in ORAUT (2017b). 

e. The Landauer Type P dosimeter was a beta-gamma-neutron dosimeter that utilized NTA film for its neutron component.

f. The LOD for this Landauer neutron dosimeter was 0.020 mrem for fast neutrons GE (1974-1980). 

Table 6-15. Whole-body neutron dosimetry missed doses for RTG Operations (1975–1991).

a

Dosimeter

type Period of use LOD (rem)

Exchange

frequency

Maximum annual

missed dose (rem)b

Landauer Type Pc Late-1975–Sep 1979 0.021 neutronsd Monthly 0.126 neutrons

Mound TLDe Oct 1979–Jun 1981 0.014 neutronsf,g Monthly 0.084 neutrons

Mound TLDe Jul 1981–Aug 1987h 0.010 neutronsf Monthly 0.060 neutrons

Landauer Type R/Ii Sep 1987h–Feb 1991 0.030 neutronsj Monthly

(1987–1989)

0.180 neutrons

Landauer Type R/Ii Sep 1987h–Feb 1991 0.030 neutronsj Quarterly

(after 1989)

0.060 neutrons (quarterly)

a. Some of the neutron dosimeters listed in this table included a component for monitoring beta-gamma exposures; however, 

only the neutron components of those dosimeters are addressed in this table. 

b. The maximum annual missed dose assignments are calculated by multiplying half of the dosimeter’s LOD value by the 

maximum number of dosimeters exchanged in a given year (NIOSH 2007).

c. The Landauer Type P dosimeter was a beta-gamma-neutron dosimeter that utilized NTA film for its neutron component.

d. The LOD for this Landauer neutron dosimeter was 0.020 mrem for fast neutrons GE (1974-1980). The LOD for this 

dosimeter has been adjusted to account for its energy response to RTG neutrons, in accordance with Section 6.3.2.4, and 

that adjustment is shown in ORAUT (2017b). 

e. The Mound TLD was a gamma-neutron dosimeter.

f. The LOD for this Mound neutron dosimeter was 0.010 mrem (ORAUT 2004). 

g. The LOD for this dosimeter has been adjusted to account for signal fading, in accordance with Section 6.3.2.3, and that 

adjustment is shown in ORAUT (2017b). 

h. During September 1987, dual Mound TLD and Landauer Type R/I neutron dosimeters were used to simultaneously monitor 

the RTG workers (Burkhart 1987b). Because it is uncertain which dosimeters were used for the reported doses, the 

dosimeter with the most favorable-to-claimant LOD values was used and the period of use date has been adjusted to reflect 

that.

i. The Landauer Type R/I dosimeter was a beta-gamma-neutron dosimeter. The neutron component to this dosimeter (i.e. 

the Type I part) utilized a TLD albedo neutron monitor with recoil device that responds to proton recoil events. 

j. The LOD for this Landauer neutron dosimeter was 0.020 mrem Landauer (1956–2002). The LOD for this dosimeter has 

been adjusted to account for its energy response to RTG neutrons, in accordance with Section 6.3.2.4, and that adjustment 

is shown in ORAUT (2017b). 
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Table 6-16. Wrist beta-gamma dosimetry missed doses. 

Dosimeter type Period of use LOD (rem)

Exchange 

frequency

Maximum annual

missed dose (rem)a

Photographic film 1957–Jun 1974 0.040 photonsb

0.040 electronsb 

Monthly 0.240 photons

0.240 electrons

Landauer Type G Jul 1974–1990 0.010 photons 

0.040 electrons

Monthly

(1974–1989)

0.060 photons

0.240 electrons

Landauer Type G Jul 1974–1990 0.010 photons 

0.040 electrons

Quarterly

(after 1989)

0.020 photons

0.080 electrons

Landauer Type K 1991–1997 0.010 photons

0.040 electrons

Quarterly 0.020 photons

0.080 electrons

a. The maximum annual missed dose assignments are calculated by multiplying half of the dosimeter’s LOD value by the 

maximum number of dosimeters exchanged in a given year (NIOSH 2007).

b. Estimated LODs for commonly used photon dosimetry (ORAUT 2006b). 

Table 6-17. Finger ring beta-gamma dosimetry missed doses. 

Dosimeter type Period of use LOD (rem)

Exchange 

frequency

Maximum annual

missed dose (rem)a

Photographic film 1957–Jun 1974 0.040 photonsb

0.040 electronsb 

Monthly 0.240 photons

0.240 electrons

Unknownc Jul 1974–Apr 1979 Unknownc Unknownc Unknownc

Landauer Type U May 1979d–1997 0.030 photons

0.040 electrons

Monthly

(1974–1989)

0.180 photons

0.240 electrons

Landauer Type U May 1979d–1997 0.030 photons

0.040 electrons

Quarterly

(after 1989)

0.030 photons

0.080 electrons

a. The maximum annual missed dose assignments are calculated by multiplying half of the dosimeter’s LOD value by the 

maximum number of dosimeters exchanged in a given year (NIOSH 2007).

b. Estimated LODs for commonly used photon dosimetry (ORAUT 2006b). 

c. Because the type of ring dosimetry used during this period was unknown at the time of this analysis, the assessment of 

any ring dosimeter results during this period will need to be handled on a case-by-case basis. Currently, there is no 

indication that ring dosimeters were ever assigned during this period. 

d. May 1979 is the earliest indication of the Landauer Type U dosimeter being used (Holliday 1980). 

6.5 UNMONITORED DOSE

6.5.1 Unmonitored Workers

The majority of the work performed at the Pinellas Plant did not involve exposures to external sources 

of radiation, which explains why a significant number of the workers were not monitored for external 

dose. Based on the review of the available dosimetry data, employees with any significant potential 

for external dose exposure appear to have been routinely monitored, as evidenced by the large 

number of monitored individuals that routinely had doses below the reporting levels. Therefore, it is 

reasonable to assume that unmonitored workers received less dose than monitored workers at the 

Pinellas Plant.

For the periods in which any Pinellas Plant worker was not monitored for external dose, an annual 

unmonitored external dose assignment of 100 mrem should be assigned. The basis for this 

unmonitored external dose assignment is in Attachment B of this document. For unmonitored periods 

that are less than 1 year in duration, the unmonitored dose assignment should be prorated, unless 

doses are being overestimated. For Plant workers who were likely exposed only to the onsite ambient 

levels of radiation, this more favorable-to-claimant unmonitored dose is assigned in lieu of the onsite 

ambient doses provided in the site’s technical basis document on occupational environmental dose 

(ORAUT 2011c), for the reasons provided in the basis for the unmonitored external dose assignment. 

In addition, the unmonitored external dose assignment should be assigned only as 100% 30- to 250-

keV photons, for the reasons provided in Attachment B.
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6.5.2 Workers Whose Dosimetry Records Are Missing

DOE has not been able to find the dosimetry records for a significant number of the Pinellas Plant 

workers. This situation was identified by comparing the DOE responses found in the claim files with 

the dosimetry records contained in ORAUT’s Site Research Database (SRDB). For a number of 

claims, DOE has indicated that no records were found, whereas NIOSH has located full or partial sets 

of dosimetry records for those claims in its SRDB. 

Because a monitored Pinellas Plant worker had a higher potential to receive external dose than an 

unmonitored worker, special considerations need to be made for workers whose dosimetry records 

are missing: 

• In the instances where sufficient records are found in the SRDB, which are identified as 

Personnel Exposure files in the claim files, no special considerations need to be made and the 

other sections of this document can be used to assess the worker’s external doses.

• In instances where there is only an indication that a worker was monitored for dose (e.g. 

records of annual whole-body doses found for some years), the worker’s job description 

should be carefully evaluated to determine if the worker was likely monitored for external dose, 

internal dose, or both. When records consisting of only annual whole-body doses are found 

for some years, the reported doses could consist of external photon doses, external neutron 

doses, internal tritium doses, or any combination of these three types of dose. If the worker 

was likely monitored for external dose, it also needs to be determined if the worker was likely 

monitored for external neutron doses and/or photon doses. Given that the available 

information for the Pinellas Plant indicates that in most instances neutron dose monitoring was 

performed in conjunction with photon dose monitoring, the default assumption is that the 

worker was monitored for both neutron doses and photon doses, when the type of dose 

monitoring is unknown. The appropriate annual missed doses should be assigned along with 

the annual unmonitored external dose assignment prescribed in Section 6.5.1, because 

missed doses were not accounted for in the unmonitored dose assignment for reasons 

indicated in Attachment B. For overestimating doses, the annual missed doses can be 

assigned based on the maximum zeros approach (the maximum dosimeters exchanged per 

year approach). For a reasonable, yet likely favorable-to-claimant estimate of a worker’s 

missed doses, the annual missed doses should be based on the maximum potential number of 

dosimeter exchanges within a year minus 1 to account for the unmonitored dose assignment 

as being a single positive dosimeter reading. 

6.6 UNCERTAINTY

When a reported external dose is based on a single measurement, the uncertainty associated with 

that individual measurement needs to be accounted for in the assigned dose. In contrast to individual 

dose measurements, the uncertainty associated with external doses based on multiple measurements 

for the same period has already been accounted for in the dose assignment. An example of this is 

unmonitored dose assignments, which are typically based on either average or upper bound annual 

doses for a population of worker doses. As a result, this section only addresses how the uncertainties 

for individual external dose measurements should be assessed.

6.6.1 Beta-Gamma Dosimeter Uncertainty

For film dosimeters, the LODs that are quoted in the literature range from about 30 to 50 mrem for 

electron and photon irradiation (Morgan 1961). These are not the expected uncertainties at larger 

electron and photon dose readings. For example, it was possible to read a photon dose of 100 mrem 

to within ±15 mrem (±15%) if the exposure involved photons with energies between several hundred 
Document No. ORAUT-TKBS-0029-6 Revision No. 02 Effective Date: 12/11/2017 Page 35 of 61

 

keV and several MeV (Morgan 1961). If the exposure involved photons with energies less than 

several hundred keV, the uncertainty was at least twice that for the more energetic photons (Morgan 

1961). Therefore, the standard error in the recorded film dosimeter doses from photons of any energy 

is estimated to be ±30%. The standard error for the recorded dose from electron irradiation was 

essentially the same as that for photon irradiation, but when an unknown mixture of electron and 

photon irradiation was involved, the standard error for the dose from beta irradiation was somewhat 

larger than ±30% (Morgan 1961). Therefore, the standard error in the recorded film dosimeter doses 

from >15 keV electrons is estimated to be ±30% for a known mixture of photon and electron 

irradiation, and higher if the mixture is unknown.

For TLDs, the uncertainty is generally lower than the uncertainty for film dosimeter results; however, 

the uncertainty is still somewhat dependent on the measured dose (NIOSH 2007). Based on that 

observation, the uncertainty associated with the Pinellas Plant’s recorded electron and photon doses 

from TLDs will be assumed to be ±30%, which is potentially favorable to claimants for some reported 

dose values.

6.6.2 Neutron Dosimeter Uncertainty

The NTA film technology used to measure neutron doses at the Pinellas Plant was similar to the 

technology used at other AEC/DOE facilities. Based on a review of the available information for such 

facilities, a reasonable uncertainty for neutron dose measurements performed using NTA film 

dosimeters is ±50%.

For the Landauer Type E neutron dosimeters, which were a polycarbonate (Lexan) dosimeter, the 

sensitivity to fast neutron radiation was reported in 1978 as 30 ± 15 mrem (±50%) (Holliday 1978). It 

was also verified that the uncertainties of the dose determinations decrease as exposures increase 

(Holliday 1978). Therefore, a reasonable uncertainty for neutron dose measurements performed 

using the Landauer Type E dosimeters is ±50%.

For the Mound TLDs, an uncertainty of ±30% should be applied based on the site-specific information 

for the Mound Site (ORAUT 2004b).

For the Landauer Type R/I neutron dosimeters, an uncertainty of ±30% should be considered 

reasonable and likely favorable to claimants, based on a review of the available information for other 

AEC/DOE facilities.

6.7 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
Document No. ORAUT-TKBS-0029-6 Revision No. 02 Effective Date: 12/11/2017 Page 36 of 61

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ORAUT (Oak Ridge Associated Universities Team), 2011c, Pinellas Plant – Occupational 

Environmental Dose, ORAUT-TKBS-0029-4, Rev. 02, Oak Ridge, Tennessee, July 15. [SRDB

Ref ID: 99867]
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ORAUT (Oak Ridge Associated Universities Team), 2011d, Use of Coworker Dosimetry Data for 

External Dose Assessment, ORAUT-OTIB-0020, Rev. 03, Oak Ridge, Tennessee, November 

14. [SRDB Ref ID: 104029]

ORAUT (Oak Ridge Associated Universities), 2011e, Pinellas Plant – Occupational Medical Dose, 

ORAUT-TKBS-0029-3, Rev. 01, Oak Ridge, Tennessee, October 13. [SRDB Ref ID: 102290] 

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Dose, ORAUT-TKBS-0008-6, Rev. 03, Oak Ridge, Tennessee, November 9. [SRDB Ref ID: 

120250]

ORAUT (Oak Ridge Associated Universities Team), 2012b, Internal Dosimetry Organ, External 

Dosimetry Organ, and IREP Model Selection by ICD-9 Code, ORAUT-OTIB-0005, Rev. 05, 

Oak Ridge, Tennessee, December 20. [SRDB Ref ID: 121336]

ORAUT (Oak Ridge Associated Universities), 2017a, MS Excel workbook titled Pinellas Plant –

Monitored Workers and Doses by Year 1960-1985, Rev. 0, Oak Ridge, Tennessee, October

10. [SRDB Ref ID: 167865] 

ORAUT (Oak Ridge Associated Universities), 2017b, MS Excel workbook titled Pinellas Plant –

Adjusted Dosimeter LODs, Rev. 0, Oak Ridge, Tennessee, October 26. [SRDB Ref ID: 

168056] 

Phillips, D., 1975, Suggestion No. 10582, memorandum to R. E. Meadors, General Electric Company, 

Neutron Devices, St. Petersburg, Florida, May 19. [SRDB Ref ID: 13125]

Poliziani, J. F., 1985, ”Assessment of 1982 Rare Gas Plant Kr-85 Doses,” memorandum to T. F. 

Pointer, Westinghouse Idaho Nuclear Company, Idaho Falls, Idaho, September 17. [SRDB 

Ref ID: 12881, pp. 14-44]

Richards, H. R., 1986, “Radiation Reports from HSR System,” memorandum to R. Greene et al., 

General Electric Company, Corporate Medical Operations, Fairfield, Connecticut, April 2. 

[SRDB Ref ID: 12243]

Schumacher, J. W., 1982, “Annual Summary of Whole Body Exposures to Ionizing Radiation,” letter to 

W. Hensley (U.S. Department of Energy, Pinellas Area Office), General Electric Company, 

Neutron Devices Department, St. Petersburg, Florida, February 9. [SRDB Ref ID: 11730]

Schumacher, J. W., 1983, “Annual Summary of Whole Body Exposures to Ionizing Radiation,” letter to 

F. A. Hoyt (U.S. Department of Energy, Pinellas Area Office), General Electric Company, 

Neutron Devices Department, St. Petersburg, Florida, February 14. [SRDB Ref ID: 11731]

Schumacher, J. W., 1984, “Annual Summary of Whole Body Exposures to Ionizing Radiation,” letter to 

F. A. Hoyt (U.S. Department of Energy, Pinellas Area Office), General Electric Company, 

Neutron Devices Department, St. Petersburg, Florida, January 25. [SRDB Ref ID: 13166, 

p. 2-12]

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Company, X-ray Department, Milwaukee, Wisconsin, February 2. [SRDB Ref ID: 12228]

Vallario, R. H., D. E. Hankins, and C. M. Unruh, 1969, AEC Workshop on Personnel Neutron 

Dosimetry, BNWL-1340, Battelle Memorial Institute, Pacific Northwest Laboratory, Richland, 

Washington. [SRDB Ref ID: 11096]
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Various authors, 1961–1990, collection of Pinellas Plant plutonium source information. [SRDB Ref ID: 

12799]

Ward, W. E., 1973, Tritium Control Technology, Technical Information Series, GEPP-104, General 

Electric Company, Neutron Devices, St. Petersburg, Florida, July 16. [SRDB Ref ID: 13128 p. 

15–40]

Ward, W. E., 1974, Account #70463, letter to D. Knuth (R. S. Landauer, Jr. & Co.), General Electric 

Company, Neutron Devices Department, St. Petersburg, Florida, June 11. [SRDB Ref ID: 

12559, p. 23–33]

Weaver, A., 1987, GEND Personnel Dosimetry Program, HP-03 Attachment 1, General Electric 

Company, Neutron Devices Department, Largo, Florida, December 7. [SRDB Ref ID: 12103, 

p. 2-4]

Weaver, A. S., 1988, External Dosimetry, HP03, Rev. 1.0, General Electric Company, Neutron 

Devices Department, St. Petersburg, Florida, October 13. [SRDB Ref ID: 12953]

Weaver, A. S., 1990, External Dosimetry, HP03, Rev, 4.0, General Electric Company, Neutron 

Devices Department, Largo, Florida, May 30. [SRDB Ref ID: 12989, p. 2-5]

Weaver, A. S., 1991, “Update on Pinellas Plant Personnel External Dosimetry Program,” 

memorandum to D. L. Gillas, General Electric Company, Neutron Devices Department, Largo, 

Florida, September 14. [SRDB Ref ID: 12111]

Weaver, A. S., 1994a, Health Physics Consensus Statement #6: Pinellas Plant Personnel External 

Dosimetry Program, Martin Marietta Specialty Components, Inc., Largo, Florida. [SRDB Ref 

ID: 12795]

Weaver, A. S., 1994b, “1993 Annual ALARA Program Report for Ionizing Radiation,” memorandum to 

C. A. Hall et al., Martin Marietta Specialty Components, Inc., Largo, Florida, March 11. [SRDB 

Ref ID: 13349, p. 7-28]

Weaver, A. S., 1996, Pinellas Plant Health Physics, Martin Marietta Specialty Components, Inc., 

Largo, Florida, February 2. [SRDB Ref ID: 12165, p. 11]

Wheeler, R. V., 1974, untitled letter to W. Ward (General Electric Company, Neutron Devices 

Department), R. S. Landauer, Jr. & Co., Glenwood, Illinois, October 10. [SRDB Ref ID: 

15848, p. 1534–1535]

Wilson, R. H., J. J. Fix, W. V. Baumgartner, and L. L. Nichols, 1990, Description and Evaluation of the 

Hanford Personnel Dosimeter Program from 1944 through 1989, PNL-7447, Battelle Memorial 

Institute, Pacific Northwest Laboratory, Richland, Washington, September. [SRDB Ref ID: 

4793]

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Landauer, Glenwood, Illinois, December 8. [SRDB Ref ID: 19707]
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GLOSSARY

absorbed dose

In external dosimetry, process in which radiation energy is imparted to material. In internal 

dosimetry, movement of material to blood regardless of mechanism.

accreditation

For external dosimetry, the assessment of whether or not a personnel dosimetry system meets 

specific criteria. The assessment includes dosimeter performance and the associated quality 

assurance and calibration programs.

accuracy

The characteristics of an analysis or determination that ensures that both the bias and 

precision of the resultant quantity will remain within the specified limits.

albedo dosimeter

Thermoluminescent dosimeter that measures the thermal, intermediate, and fast neutrons 

scattered and moderated by the body or a phantom from an incident fast neutron flux.

alpha radiation

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2.

backscatter

Reflection or refraction of radiation at angles over 90 degrees from its original direction.

beta particle

See beta radiation.

beta radiation

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 that of 

a proton. A negatively charged beta particle is identical to an electron. A positively charged 

beta particle is a positron.

curie (Ci)

Traditional unit of radioactivity equal to 37 billion (3.7 × 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

densitometer

Instrument that uses a photoelectric cell to measure the transition of light through developed 

X-ray film to determine its optical density.

DOE Laboratory Accreditation Program (DOELAP)

Program for accreditation by DOE of DOE site personnel dosimetry and radiobioassay 

programs based on performance testing and the evaluation of associated quality assurance, 

records, and calibration programs.

dose

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of roentgens, rads, reps, or grays.
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dose equivalent

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose.

dose equivalent index

Historical measure for neutron source calibration defined by the International Commission on 

Radiation Units and Measurements as the sum of the maximum dose equivalents delivered 

within a sphere at any depth for the respective neutron energies even though the maximum 

dose occurred at different depths and discounting the outer 0.07-millimeter-thick shell. Also 

called unrestricted dose equivalent index.

dosimeter

Device that measures the quantity of received radiation, usually a holder with radiationabsorbing filters and radiation-sensitive inserts packaged to provide a record of absorbed dose 

received by an individual. See albedo dosimeter, film dosimeter, neutron film dosimeter, 

pocket ionization chamber, thermoluminescent dosimeter, and track-etch dosimeter.

dosimetry system

System for assessment of received radiation dose. This includes the fabrication, assignment, 

and processing of external dosimeters, and/or the collection and analysis of bioassay samples, 

and the interpretation and documentation of the results.

electron radiation

See beta radiation.

error

Difference between the correct, true, or conventionally accepted value and the measured or 

estimated value. Sometimes used to mean estimated uncertainty. See accuracy and 

uncertainty.

exchange period (frequency)

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

exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

extremities

The arms from and including the elbow through the fingertips and the legs from and including 

the knee and patella through the toes.

fast neutron

Neutron with energy equal to or greater than 10 kiloelectron-volts. This type of neutron causes 

fission in some isotopes (e.g., 238U, 239Pu). See intermediate neutron and slow neutron.

favorable to claimants

In relation to dose reconstruction for probability of causation analysis, having the property of 

ensuring that there is no underestimation of potential dose. This often means the assumption 

of a value that indicates a higher dose than is likely to have actually occurred in the absence of 

more accurate information. See probability of causation.
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film

In the context of external dosimetry, radiation-sensitive photographic film in a light-tight 

wrapping. See film dosimeter. 

film dosimeter

Package of film for measurement of ionizing radiation exposure for personnel monitoring 

purposes. A film dosimeter can contain two or three films of different sensitivities, and it can 

contain one or more filters that shield parts of the film from certain types of radiation. When 

developed, the film has an image caused by radiation measurable with an optical 

densitometer. Also called film badge.

filter

Material used in a dosimeter to adjust radiation response to provide an improved tissue 

equivalent or dose response.

gamma radiation

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are identical 

to X-ray photons of high energy; the difference is that X-rays do not originate in the nucleus. 

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. See 

alpha radiation, beta radiation, gamma radiation, neutron radiation, photon radiation, and X-ray 

radiation. 

isotope

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties.

kiloelectron-volt (keV)

Unit of particle energy equal to 1,000 (1 × 103

) electron-volts.

luminescence

Emission of light from a material as a result of some excitation. See thermoluminescence. 

limit of detection (LOD)

Minimum level at which a particular device can detect and quantify exposure or radiation. Also 

called lower limit of detection and detection limit or level. See minimum detectable level. 

megaelectron-volt (MeV)

Unit of particle energy equal to 1 million (1 × 106

) electron-volts.

monitoring

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 
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monitoring includes internal and external dosimetry including interpretation of the 

measurements.

neutron (n)

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

neutron film dosimeter

Film dosimeter with a nuclear track emulsion, type A, film packet.

neutron radiation

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons. See neutron. 

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz.

photon radiation

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays.

probability of causation (POC)

For purposes of dose reconstruction for the Energy Employees Occupational Illness 

Compensation Program Act of 2000, the percent likelihood, at the 99th percentile, that a 

worker incurred a particular cancer from occupational exposure to radiation.

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joules per kilogram) of absorbing tissue. The rad has 

been replaced by the gray in the International System of Units (100 rad = 1 gray). The word 

derives from radiation absorbed dose.

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer. See ionizing radiation. 

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei. 

radioisotopic thermoelectric generator (RTG)

Generator that obtains its power from passive (natural) radioactive decay using thermocouples 

to convert the heat of decay into electricity.

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 
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factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

roentgen

Unit of photon (gamma or X-ray) exposure for which the resultant ionization liberates a positive 

or negative charge equal to 2.58 × 10-4 coulombs per kilogram (or 1 electrostatic unit of 

electricity per cubic centimeter) of dry air at 0 degrees Celsius and standard atmospheric 

pressure. An exposure of 1 roentgen is approximately equivalent to an absorbed dose of 

1 rad in soft tissue for higher energy photons (generally greater than 100 kiloelectron-volts).

scattering

Change in direction of radiation by refraction or reflection, often accompanied by a decrease in 

radiation due to absorption by the refracting or reflecting material.

shallow dose equivalent

Dose equivalent in units of rem or sievert at a depth of 0.07 millimeters (7 milligrams per 

square centimeter) in tissue equal to the sum of the penetrating and nonpenetrating doses.

shielding

Material or obstruction that absorbs ionizing radiation and tends to protect personnel or 

materials from its effects.

skin dose

See shallow dose equivalent.

thermal neutron

Neutron in thermal equilibrium with its surroundings having an average energy of 

0.025 electron-volts.

thermoluminescence

Property that causes a material to emit light as a result of heat.

thermoluminescent dosimeter (TLD)

Device for measuring radiation dose that consists of a holder containing solid chips of material 

that, when heated, release the stored energy as light. The measurement of this light provides 

a measurement of absorbed dose.

U.S. Atomic Energy Commission (AEC)

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The U.S. Energy Research and Development 

Administration and the U.S. Nuclear Regulatory Commission assumed separate duties from 

the AEC in 1974. The U.S. Department of Energy succeeded the U.S. Energy Research and 

Development Administration in 1979.

uncertainty

Standard deviation of the mean of a set of measurements. The standard error reduces to the 

standard deviation of the measurement when there is only one determination. See accuracy

and error. Also called standard error.

whole-body dose

Dose to the entire body excluding the contents of the gastrointestinal tract, urinary bladder, 

and gall bladder and commonly defined as the absorbed dose at a tissue depth of 
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10 millimeters (1,000 milligrams per square centimeter). Also called penetrating dose. 

See dose. 

X-ray

See X-ray radiation. 

X-ray radiation

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.
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ATTACHMENT A 

EXAMPLES OF PINELLAS PLANT EXTERNAL DOSIMETRY RECORDS

LIST OF FIGURES

FIGURE TITLE PAGE

A-1 Example of dosimetry records used during 1957–1959.......................................................... 51

A-2 Example of dosimetry records used during 1959–1965.......................................................... 52

A-3 Example of dosimetry records used during 1964–1970.......................................................... 52

A-4 Example of dosimetry records used during 1971–1980.......................................................... 53

A-5 Example of dosimetry records used during 1980–1987.......................................................... 54

A-6 Example of dosimetry records used during 1988–1997.......................................................... 54

Attachment A provides examples of the Pinellas Plant’s external dosimetry records that contain 

individual dosimeter results. These examples were obtained from EEOICPA claim records in the 

NOCTS database. All personal information in these examples has been blacked out. Note that some 

of the records formats used throughout the Pinellas Plant’s history overlapped each other.
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ATTACHMENT A 

EXAMPLES OF PINELLAS PLANT EXTERNAL DOSIMETRY RECORDS (continued)

Figure A-1. Example of dosimetry records used during 1957–1959. 
Document No. ORAUT-TKBS-0029-6 Revision No. 02 Effective Date: 12/11/2017 Page 52 of 61

ATTACHMENT A 

EXAMPLES OF PINELLAS PLANT EXTERNAL DOSIMETRY RECORDS (continued)

Figure A-2. Example of dosimetry records used during 1959–1965. 

Figure A-3. Example of dosimetry records used during 1964–1970. 
Document No. ORAUT-TKBS-0029-6 Revision No. 02 Effective Date: 12/11/2017 Page 53 of 61

ATTACHMENT A 

EXAMPLES OF PINELLAS PLANT EXTERNAL DOSIMETRY RECORDS (continued)

Figure A-4. Example of dosimetry records used during 1971–1980. 
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ATTACHMENT A 

EXAMPLES OF PINELLAS PLANT EXTERNAL DOSIMETRY RECORDS (continued)

Figure A-5. Example of dosimetry records used during 1980–1987. 

Figure A-6. Example of dosimetry records used during 1988–1997. 
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ATTACHMENT B 

BASIS FOR UNMONITORED EXTERNAL DOSE ASSIGNMENT

TABLE OF CONTENTS

SECTION TITLE PAGE

B.1 Background............................................................................................................................ 56

B.2 Basis for Upper-95th Percentile Dose .................................................................................... 56

B.3 Comparison of Maximum Likelihood Doses to Unmonitored Dose Assignment...................... 58

B.4 Inapplicability of ORAUT-OTIB-0020...................................................................................... 59

B.5 Assigning Unmonitored Doses as Photon Versus Neutron..................................................... 59

LIST OF TABLES

TABLE TITLE PAGE

B-1 Evaluation of Pinellas Plant doses available in the SRDB...................................................... 57

B-2 Evaluation of POCs associated with assigning unmonitored doses as either photon or 

neutron dose..........................................................................................................................61
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ATTACHMENT B 

BASIS FOR UNMONITORED EXTERNAL DOSE ASSIGNMENT (continued)

B.1 BACKGROUND

Based on an evaluation of the available dosimetry records for the Pinellas Plant, personnel who had 

any potential to receive significant external dose were likely monitored. In addition, the majority of the 

work performed at the Pinellas Plant was nonradiological, which explains why a significant number of 

the employees were not monitored. However, there was the potential to receive incidental external 

doses that were small (i.e., < 20 mrem/yr) but were much more significant than the highest onsite 

ambient doses for the plant [i.e., >> 3.66 × 10-13 mrem/yr, from the Plant’s environmental dose TBD 

(ORAUT 2011c)]. One of the most common types of incidental radiation exposures was attributable 

to Plant tours, which included visits to the neutron generator testing areas. Based on information 

provided in the claimant telephone interviews, personnel on these tours were allowed to witness the 

neutron generators being tested without being provided with dosimetry. Pinellas Plant health physics 

records indicate that doses at the 3-ft exclusion boundary for the neutron generator tests were 

9.9 mrem per test (Weaver 1996). Assuming that personnel on a Plant tour would not have likely 

stayed in the neutron generator testing area for more than two tests, these exposed personnel might 

have received an incidental dose of almost 20 mrem.

Because the incidental doses at the Pinellas Plant were likely much larger than the onsite ambient 

doses, it is not appropriate to assign onsite ambient doses to all unmonitored workers unless it can be 

established that there was no potential for incidental dose due to an employee’s job. Because a 

determination that incidental doses were not received can rarely be defended, unmonitored external 

doses should be assigned for all workers in lieu of the onsite ambient doses for the Plant.

B.2 BASIS FOR UPPER-95TH PERCENTILE DOSE

A review of the available dosimetry data for the Pinellas Plant indicates that on average 95% of the 

monitored workers at the plant received annual doses ≤100 mrem. This determination is based on an 

evaluation of whole-body dose information that was available in the following documents: GE 1956–

1980, 1957–1990, 1975; Greene 1985b, 1986; Holliday 1977, 1979, 1981, 1983; LMSC 1996; 

Schumacher 1982, 1983, 1984; and Weaver 1994b.

Whole-body dose information was used for the evaluation because it was the only form of data that 

was consistently available throughout the years the Pinellas Plant operated. Unfortunately, the 

Pinellas Plant’s reported whole-body doses consist of external photon, external neutron, and internal 

tritium doses, and the contribution attributable to each of these components cannot be determined for 

most years. In addition, the doses evaluated represent population doses that do not account for 

potential missed doses. Table B-1 provides summarized results of the dosimetry data that were 

evaluated. The average percentage of annual whole-body doses below 100 mrem was 94.7% (95% 

when rounded to the nearest percent), for the years that were evaluated. However, for a select 

number of years, more detailed dose information was provided, and it was determined that a 

significant number of the annual whole-body doses were either entirely or mostly attributable to tritium 

dose. This is consistent with the types of operations performed at the Pinellas Plant. Therefore, the 

upper-95th-percentile external dose for monitored workers is likely to be less than 100 mrem/yr.
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ATTACHMENT B 

BASIS FOR UNMONITORED EXTERNAL DOSE ASSIGNMENT (continued)

Table B-1. Evaluation of Pinellas Plant doses available in the SRDB. 

Year

Total number of 

annual doses 

checked

Number of 

annual doses 

≥100 mrem

Percentage of 

annual doses 

below 100 mrem

Number of 

annual doses 

≤20 mrem

Percentage of 

annual doses 

≤20 mrem

1957 71 12 83.1 38 53.5

1958 142 4 97.2 102 71.8

1959 205 43 79.0 114 55.6

1960 232 40 82.8 147 63.4

1961 265 39 85.3 168 63.4

1962 259 36 86.1 159 61.4

1963 258 20 92.2 197 76.4

1964 292 13 95.5 212 72.6

1965 280 37 86.8 233 83.2

1966 320 17 94.7 277 86.6

1967 351 8 97.7 328 93.4

1968 355 3 99.2 340 95.8

1969 344 9 97.4 326 94.8

1970 293 9 96.9 258 88.1

1971 282 6 97.9 261 92.6

1972 292 7 97.6 267 91.4

1973 280 10 96.4 254 90.7

1974 303 8 97.4 265 87.5

1975 276 2 99.3 246 89.1

1976 255 5 98.0 195 76.5

1977 246 27 89.0 159 64.6

1978 262 20 92.4 176 67.2

1979 313 17 94.6 244 78.0

1980 383 23 94.0 249 65.0

1981 388 18 95.4 324 83.5

1982 418 4 99.0 286 68.4

1983 396 11 97.2 263 66.4

1984 428 13 97.0 296 69.2

1985 395 15 96.2 253 64.1

1986 322 Not applicable Not applicable Not applicable Not applicable

1987 248 Not applicable Not applicable Not applicable Not applicable

1988 263 4 98.5 245 93.2

1989 285 5 98.2 270 94.7

1990 284 4 98.6 258 90.8

1991 288 1 99.7 274 95.1

1992 280 0 100.0 275 98.2

1993 245 0 100.0 242 98.8

1994 NA Not applicable Not applicable Not applicable Not applicable

1995 215 0 100.0 Not applicable Not applicable

1996 Not applicable Not applicable Not applicable Not applicable Not applicable

1997 Not applicable Not applicable Not applicable Not applicable Not applicable

1998 Not applicable Not applicable Not applicable Not applicable Not applicable

Averages Not applicable Not applicable 94.7 Not applicable 79.6
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ATTACHMENT B

BASIS FOR UNMONITORED EXTERNAL DOSE ASSIGNMENT (continued)

Upon review of the dosimetry data in the referenced documents listed above and in Table 6-1 of this 

document, many discrepancies regarding the numbers of individuals monitored were noted. 

Therefore, it is not recommended that the 95th-percentile doses for individual years be tabulated 

unless a better data set is located. By averaging the data over several years, the potential for 

incurring a significant error for a given year would be minimized. In addition, including the dose 

contribution attributable to tritium helps to compensate for any potential errors and helps to ensure 

that the 95th-percentile dose is favorable to claimants. Because the majority (>79% each year on 

average) of the annual whole-body doses reported for Pinellas Plant workers were ≤20 mrem, which 

is an individual dosimeter’s limit of detection for some of the dosimeters used at the Plant, it appears 

that the Plant monitored workers if they had any potential to receive dose. As a result, it is unlikely 

that any unmonitored workers would have received doses approaching the upper-95th percentile 

doses. Therefore, it is very unlikely that any errors in the 95th-percentile dose calculation attributable 

to these discrepancies would result in an underestimate of an individual’s actual unmonitored dose.

B.3 COMPARISON OF MAXIMUM LIKELIHOOD DOSES TO UNMONITORED DOSE 

ASSIGNMENT

Area dosimeter results with their placement locations identified in relation to the external radiation 

source terms are very limited. However, detailed area dosimeter results were available for the period 

from 1984 through June 1986 in GE (1984–1986).

The area dosimeter results available for Areas 128, 131, and 183 were evaluated. These areas 

represent photon and neutron exposure areas, which are in the neutron generator operational areas 

of the Plant. Area 131 was the final testing area of the neutron generators. Area 128 was 

represented by dosimeter locations 9000, 9001, 9003, 9004, 9008, and 9019. Area 131 was 

represented by dosimeter locations 9026 and 9038. Area 183 was represented by dosimeter location 

9020. All of these dosimeter locations were reported as being within 24 to 65 in. of the closest 

radiation source. For the period from 1984 through June 1986, only a single photon dosimeter result 

above the dosimeter’s LOD was reported and only two neutron dosimeter results at two different 

locations were reported as being above the dosimeter’s LOD.

Based on the area dosimeter data that were evaluated, the worst-case annual photon dose is 

20 mrem plus 11 dosimeters worth of missed dose, and the worst-case annual neutron dose is 

40 mrem plus 11 dosimeters worth of missed dose. Because area dosimeter results correspond to a 

8,760-hour year versus a 2,500-hour work-year, the worst-case annual doses needed to be adjusted 

by a maximized occupancy factor of 0.2854 (2,500/8,760). For the photon and neutron dosimeters 

used during this period, no other dose adjustments were needed as indicated in Section 6.3. For 

1984–1986, the LOD/2 values for the applicable photon and neutron dosimeters were 5 mrem and 10 

mrem per dosimeter exchange, respectively. Based on those values, the maximum likely external 

dose to an unmonitored worker would be <64 mrem/yr (21 mrem/yr from photons and 43 mrem/yr 

from neutrons). Because it is unlikely that a Pinellas Plant worker would have spent 100% of their 

time within 65 in. of a radiation source without being monitored, this is likely a significant overestimate 

of an unmonitored worker’s potential external dose.

The potential doses to workers in the RTG areas of Building 400 are potentially much higher than 

64 mrem/yr. However, because all types of plutonium are considered to be special nuclear material, 

access to the RTG areas was more strictly limited. Because Pinellas Plant personnel having work 

assignments in radiation areas were required to wear dosimeters per General Operating Procedures 

(Greene 1985b; GE 1982), and because access to the RTG areas would have been controlled for 

security purposes, it is much less likely that personnel would have been able to receive significant 
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BASIS FOR UNMONITORED EXTERNAL DOSE ASSIGNMENT (continued)

unmonitored doses in the RTG areas. Therefore, the potential doses for unmonitored workers in the 

RTG areas were also likely less than 64 mrem/yr.

B.4 INAPPLICABILITY OF ORAUT-OTIB-0020

Because an unmonitored worker’s potential to receive external dose at the Pinellas Plant was 

significantly less than the dosimetry program’s ability to detect doses that low, the unmonitored dose 

approach described in ORAUT-OTIB-0020, Use of Coworker Dosimetry Data for External Dose 

Assessment (ORAUT 2011d), which uses only coworker data, can result in an unreasonable 

overestimate of an unmonitored Pinellas Plant worker’s external doses. This determination is 

supported by an analysis of the available area dosimeter results, which have a longer effective 

monitoring period per dosimeter exchange, and the fact that on average >79% of the annual wholebody doses reported for Plant workers were ≤20 mrem.

The ORAUT-OTIB-0020 approach requires the LOD/2 approach for accounting for the coworker’s 

potential missed doses, which is favorable to claimants (ORAUT 2011d). Based on the maximum 

likelihood doses above, the ORAUT-OTIB-0020 approach results in an unreasonably excessive 

unmonitored dose estimate for unmonitored Pinellas Plant workers. Because of this and because the 

95th-percentile dose is already greater than the maximum likelihood dose, the potential missed doses 

were not factored into the unmonitored dose assignment for the Pinellas Plant. Another alternative 

would be to use the maximum likelihood dose of 64 mrem/yr as the unmonitored dose assignment. 

However, a decision was made to use the higher unmonitored dose assignment of 100 mrem/yr to 

ensure that the worker’s external doses are not underestimated.

B.5 ASSIGNING UNMONITORED DOSES AS PHOTON VERSUS NEUTRON

Because the unmonitored external dose assignment includes both photon and neutron dose 

components and because the proportions of those dose components are unknown and highly 

variable, an evaluation was performed to determine which of the applicable radiation types and 

energy groups to assign the unmonitored doses as.

The only photon energy distribution that is used for the external doses for Pinellas Plant workers is 

100% 30-250 keV photons, which is also the most favorable-to-claimant photon energy distribution. 

Therefore, only 30-250 keV photons were evaluated for the potential unmonitored dose assignment.

Based on the information in the main body of this document, there are only two neutron energy 

distributions for the various areas at the Pinellas Plant. For the neutron generators areas at the Plant, 

the neutron energy distribution was 100% 2-20 MeV neutrons. For the RTG areas in Building 400, the 

neutron energy distribution was 50% 0.1-2 MeV and 50% 2-20 MeV neutrons. Given that it was 

unlikely that a worker would have been able to spend any significant time in the presence of the PuO2

sources in the RTG areas without having to wear a dosimeter, the predominant neutron energies that 

unmonitored workers at the Plant were likely exposed to were 2-20 MeV neutrons. Therefore, only 

2-20 MeV neutrons were evaluated for the potential unmonitored dose assignment.

The unmonitored dose assignment of 100 mrem/yr was assigned as photon dose and then neutron 

dose to compare the POC results using each of the IREP cancer models. Because ORAUT-OTIB0005, Internal Dosimetry Organ, External Dosimetry Organ, and IREP Model Selection by ICD-9 Code

(ORAUT 2012b), indicates that there could be multiple external organ dose selections for a given 

IREP cancer model, this evaluation was performed using only the external organ dose selection that 

resulted in the highest dose (i.e., the selection with the highest organ dose conversion factor) for a 
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BASIS FOR UNMONITORED EXTERNAL DOSE ASSIGNMENT (continued)

given IREP cancer model selection. Each IREP cancer model was evaluated using the following 

hypothetical scenario assumptions:

1. The hypothetical worker was a male, with the exception of cancer models that are only 

applicable to females.

2. The worker was 18 years old at the time of the first exposure. Therefore, a year of birth of 

1939 and a first year of employment date of 1957 were selected.

3. The worker was assumed to be employed and unmonitored for 5 years (i.e., during all of 1957 

through 1961).

4. Only an unmonitored dose of 100 mrem was assigned for each year of employment.

5. The date of diagnosis was set at 12 years after the last year of employment (i.e., 1973).

Because the unmonitored dose assignment represents a population dose, no uncertainty correction 

factors were applied to the unmonitored dose assignment. As indicated in the main body of this 

document, no dosimeter correction factors need to be applied to the photon doses. Even though the 

main body of this document indicates that dosimeter corrections factors would be applied to the 

reported neutron doses for most years, no dosimeter correction factors were applied when the 

unmonitored doses were evaluated as neutron doses. The basis for this is that the unmonitored dose 

assignment is essentially an arbitrary value that has been determined to be more favorable to 

claimants than the maximum likelihood dose for the Pinellas Plant, and the evaluation of the 

maximum likelihood dose for the Plant utilized neutron dosimeter data that did not require any 

adjustments, as indicated in Section 6.3. In addition, the ICRP Publication 60 correction factor for 

2-20 MeV neutron doses is 1.00 for the Plant, so this correction factor had no impact on the neutron 

doses being evaluated (ICRP 1991).

The results of this evaluation are summarized in Table B-2 below. With the only exceptions being the 

acute myeloid leukemia, chronic myeloid leukemia, and leukemia (less chronic lymphocytic leukemia) 

IREP cancer models, the assignment of the unmonitored doses as 100% 30-250 keV photons 

provides a more favorable-to-claimant POC. At the time of this document’s preparation, it was 

determined that out of 393 Pinellas Plant claims, which excludes pulled claims, there are only 5 

leukemia claims that are affected by the photon-versus-neutron unmonitored dose issue. However, 

because the unmonitored workers at any site only need to be assigned 50th-percentile coworker 

doses and because the unmonitored external dose assignment is greater than the maximum 

likelihood dose, assigning an unmonitored external dose of 100 mrem/yr as only photon dose is still 

considered to be favorable to claimants for the leukemia cases based on the results of this evaluation.
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BASIS FOR UNMONITORED EXTERNAL DOSE ASSIGNMENT (continued)

Table B-2. Evaluation of POCs associated with assigning unmonitored doses as either 

photon or neutron dose. 

Applicable cancer models

Worst-case 

external dose 

organ selection

Photon 

dose POC 

(%)

Neutron 

dose POC 

(%)

Acute lymphocytic leukemia Red bone marrow 17.89 11.01

Acute myeloid leukemia Red bone marrow 10.59 11.83

All digestive Stomach 4.34 2.56

All male genitalia Testes 3.57 2.21

Bladder Bladder 4.68 3.52

Bone Bone surfaces 4.98 3.38

Breast Breast (female) 7.58 4.04

Chronic myeloid leukemia Red bone marrow 12.63 12.71

Colon Colon 5.33 3.96

Connective tissue Thyroid 5.36 3.45

Esophagus Esophagus 4.95 3.99

Eye Eye 4.72 3.08

Female genitalia (less ovary) Bladder 0.04 0.03

Gallbladder Bladder 11.66 6.65

Leukemia (less CLL) Red bone marrow 15.69 18.34

Liver Liver 17.14 10.61

Lung Thyroid 5.72 3.03

Lymphoma and multiple myeloma Thyroid 3.43 1.77

Malignant melanoma Skin 10.09 7.63

Nervous system Thyroid 2.41 1.45

Non-melanoma - basal cell carcinoma Skin 10.20 7.82

Non-melanoma - squamous cell carcinoma Skin 0.33 0.15

Oral cavity and pharynx Thyroid 2.59 1.54

Other and ill-defined sites Thyroid 5.19 3.13

Other endocrine Thyroid 6.30 4.00

Other respiratory Thyroid 2.26 1.05

Ovary Bladder 5.42 3.84

Pancreas Stomach 2.47 1.60

Rectum Colon 1.70 1.17

Stomach Stomach 7.95 4.81

Thyroid Thyroid 16.08 7.10

Urinary organs (less bladder) Testes 6.24 4.24