ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 42

Document Title:

Pinellas Plant – Occupational External 

Dose

Document Number: ORAUT-TKBS-0029-6 

Revision: 00 PC-1 

Effective Date: 08/03/2006

Type of Document: TBD

Supersedes: None

Subject Experts: Donald. E. Palmrose, Milton E. Gorden, and Paul J. Demopoulos

Document Owner

Approval: Signature on File Approval Date: 09/13/2005

Mark D. Notich, TBD Team Leader

Approval: Signature on File Approval Date: 09/13/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 09/13/2005

Richard E. Toohey, Project Director

Approval: Signature on File Approval Date: 09/15/2005

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0029-6 Revision No. 00 PC-1 Effective Date: 08/03/2006 Page 2 of 42

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

01/07/2005 00-A New Technical Basis Document for Pinellas – Site External Dose. 

Initiated by Mark D. Notich.

06/20/2005 00-B Incorporates Task 5 review comments. Initiated by Mark D. Notich.

08/15/2005 00-C Incorporates OCAS review comments. Initiated by Mark D. Notich.

08/22/2005 00-D Incorporates additional OCAS review comments. Initiated by Mark 

D. Notich.

09/07/2005 00-E Incorporates further OCAS review comments and adds Paul 

Demopoulos to list of subject experts. Initiated by Mark D. Notich.

09/15/2005 00 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_______________________________ Paul J. Demopoulos, TBD Team Leader07/25/2006

Signature on File_______________________________ John M. Byrne, Task 3 Manager07/25/2006

Signature on File_______________________________ Edward F. Maher, Task 5 Manager07/24/2006

Signature on File_______________________________ Kate Kimpan, Projector Director07/27/2006

Signature on File_______________________________ James W. Neton, Associate Director for Science08/03/2006
Document No. ORAUT-TKBS-0029-6 Revision No. 00 Effective Date: 09/15/2005 Page 3 of 42

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

6.1 Introduction................................................................................................................................. 6

6.1.1 Purpose...........................................................................................................................7

6.1.2 Scope..............................................................................................................................7

6.1.3 Sources of Radiation .......................................................................................................8

6.1.4 Occupational Dosimetry Program Overview ..................................................................10

6.2 Dose Reconstruction Parameters ............................................................................................. 11

6.2.1 Historical Administrative Practices.................................................................................11

6.2.1.1 Performance Testing....................................................................................... 14

6.2.1.2 Reports ........................................................................................................... 14

6.2.2 Dosimetry Technology...................................................................................................14

6.2.3 Dosimetry Calibration Practices.....................................................................................15

6.2.4 Workplace Radiation Fields...........................................................................................17

6.2.4.1 Beta-Gamma Exposures................................................................................. 19

6.2.4.2 Neutron Exposures ......................................................................................... 20

6.3 Adjustments to Recorded Dose................................................................................................. 21

6.3.1 Photon Dose Adjustments .............................................................................................23

6.3.2 Neutron Dose Adjustments............................................................................................23

6.3.3 Beta Dose Adjustments ................................................................................................. 24

6.4 Missed Dose .............................................................................................................................24

6.4.1 Photon and Beta Missed Dose ......................................................................................24

6.4.1.1 Missed Dose by Facility or Location................................................................ 25

6.4.1.2 Missed Dose by Dosimeter Type and Year ..................................................... 26

6.4.1.3 Missed Dose by Energy Range....................................................................... 26

6.4.2 Neutron Missed Dose....................................................................................................26

6.4.2.1 Missed Dose by Dosimeter Type and Year ..................................................... 27

6.5 Uncertainty................................................................................................................................28

References ........................................................................................................................................29

Glossary ............................................................................................................................................33

Attachment A Examples of Pinellas Plant Exposure Records ............................................................ 38
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LIST OF TABLES

TABLE TITLE PAGE

6-1 X-ray producing equipment ......................................................................................................9

6-2 Personnel radiation dosimetry from AEC annual reports ........................................................ 10

6-3 Pinellas Plant historical dosimetry events............................................................................... 12

6-4 Dosimetry used at Pinellas Plant for external whole-body, wrist and extremity 

exposures ..............................................................................................................................16

6-5 Minimum Detectable Level (MDL) and Maximum Potential Missed Photon or 

Beta Dose ..............................................................................................................................17

6-6 Radionuclides contributing to workplace radiation fields......................................................... 18

6-7 Beta and photon radiation energies and percentages for Pinellas Plant 

processes ..............................................................................................................................19

6-8 Selection of neutron radiation energies for Pinellas Plant processes...................................... 20

6-9 Internal and external doses among Pinellas claimants (mrem)............................................... 22

6-10 Selection of neutron radiation energies and associated dose fractions for the 

RTG building 400 Pinellas Plant processes ............................................................................ 24

6-11 Photon and beta dosimetry missed dose by facility or location............................................... 25

6-12 Neutron dosimetry missed dose by facility or location ............................................................ 27

LIST OF FIGURES

FIGURE TITLE PAGE

6-1 Regenerated example of a Landauer dosimetry report from the 1970s .................................. 13

6-2 RTG Plutonium Source Neutron Energy Spectrum – PuO2 Microspheres .............................. 21
Document No. ORAUT-TKBS-0029-6 Revision No. 00 Effective Date: 09/15/2005 Page 5 of 42

ACRONYMS AND ABBREVIATIONS

AEC U.S. Atomic Energy Commission

ALARA as low as reasonably achievable

ANSI American National Standards Institute

CFR Code of Federal Regulations

DOE U.S. Department of Energy

DOELAP U.S. Department of Energy’s Laboratory Accreditation Program

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

GEND General Electric Aerospace, Neutron Devices

HSR General Electric Health and Safety Record system

MDL Minimum Detection Level

NIOSH National Institute for Occupational Safety and Health

NOCTS NIOSH Office of Compensation Analysis and Support (OCAS) Claims Tracking System

NTA nuclear emulsion type A film neutron dosimeter

OCAS NIOSH Office of Compensation Analysis and Support

QA Quality Assurance

RBE relative biological effectiveness

RGD radiation generating device

RTG radioisotopic thermoelectric generator

TBD technical basis document

TLD thermoluminescent dosimeter

U.S.C. United States Code

XRE X-ray emission units

XRD X-ray diffraction units
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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 historic background information and guidance to assist in the preparation of 

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

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

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

reconstruction.

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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

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

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

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

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

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

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

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

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

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

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

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

derived:

] guidelines established under 

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

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

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

 

1 The U.S. Department of Labor is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0029-6 Revision No. 00 PC-1 Effective Date: 08/03/2006 Page 7 of 42

• Radiation from naturally occurring radon present in conventional structures

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

This Site Profile provides specific information concerning documentation of historical practices at the 

Pinellas Plant.

General Electric Aerospace, Neutron Devices (GEND) operated the Pinellas Plant for the U.S. 

Department of Energy (DOE) from its initial startup in January 1957 until 1992, when Lockheed Martin 

Specialty Components, Inc. took over until nuclear operations ended in 1994; some clean-up 

operations occurred until the plant closure in 1997. The Plant was built to manufacture neutron 

generators, a principal component in nuclear weapons. The neutron generators consisted of a 

miniaturized linear ion accelerator assembled with pulsed electric power supplies. The ion accelerator, 

or neutron tube, required ultra-clean, 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 operations. It later 

manufactured other products including neutron detectors, radioisotopic thermoelectric generators 

(RTGs), high-vacuum switch tubes, specialty capacitors, and specialty batteries (DOE 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 (DOE 1996). 

6.1.1

This technical basis document (TBD) represents a specific support mechanism concerning 

documentation of external dosimetry historical practices at the Pinellas Plant. This external dose TBD 

can be used to evaluate external dosimetry data for monitored workers and can serve as a 

supplement to individual monitoring data. For unmonitored workers, information is present in this 

document that will provide for estimations of external doses. This document provides a site profile of 

Pinellas that contains technical basis information to be used to evaluate the total occupational 

external radiation dose for EEOICPA claimants.

Purpose

6.1.2

This document provides supporting technical data to evaluate the total Pinellas occupational radiation 

dose that may reasonably be associated with the worker’s radiation exposure. This dose results from 

exposure to external radiation sources in Pinellas facilities that would be added to Pinellas

occupationally-required diagnostic X-ray examinations, and to on-site environmental releases, if 

applicable, in order to determine the total external dose. Also included are techniques to estimate the 

dose that may have occurred while an employee was not monitored, inadequately monitored, dose 

that may have been missed due to analytical detection limits, or whose monitoring records are 

incomplete or missing (i.e., missed dose). Over the years new and more reliable scientific methods 

and protection measures have been deployed. The methods needed to account for these changes 

are also identified in this document.

Scope

This Occupational External Dosimetry TBD describes the external dosimetry program at Pinellas. 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 present in the work environment. 

It also discusses the specific details of the evaluation of doses measured from exposure to beta, 

gamma, and neutron radiation; sources of bias, workplace radiation field characteristics, responses to 

different beta/gamma and neutron dosimeters in the workplace fields, and adjustments to the 

recorded dose measured by these dosimeters during specific years.
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6.1.3

The manufacture of Pinellas Plant products required the use of radioisotopes. These included tritium, 

depleted uranium, 85Kr, 238Pu, and 239Pu. In addition, the Plant used various analytical devices and 

calibration sources as part of the production process or for radiation detection device maintenance 

and calibration.

Sources of Radiation

The Pinellas Plant stored tritium gas in beds of depleted uranium metal. The uranium and tritium 

joined to form uranium tritide, which permits the storage of large quantities of ultra-pure tritium in a 

small volume. A search of Plant records indicated that, from the start of operations in 1957 until 

November 1993, the total amount of tritium received was 234.1 g (8.3 oz), or 8.14 × 1016 Bq (2.2 

million Ci). The total amount of tritium shipped as product, waste, or material returned to DOE was 

164.6 g (5.8 oz), or 5.92 × 1016 Bq (1.6 million Ci). The amount calculated to have decayed away was 

58.4 g (2 oz), or 2.22 × 1016 Bq (0.6 million Ci), and the amount released into wastewater or air was 

11.1 g (0.4 oz), or 3.7 × 1015 Bq (0.1 million Ci) (GEND 1993). The majority of the atmospheric 

releases of tritium were through the four exhaust stacks on Buildings 100, 200, and 800. Tritium is 

primarily a source of internal exposure through inhalation and ingestion, but not of external exposure 

because it is a low-energy beta emitter that is readily absorbed in the dead layer of the skin.

Depleted uranium metal in sealed containers stored tritium at normal pressure and temperature. 

Heating can decompose the tritium bonds in uranium tritide to release ultra-pure tritium while the 

remaining depleted uranium metal is ready for fast and reversible uptake of tritium at lower 

temperatures. The depleted uranium metal was fully contained in the tritium storage vessel. The 

storage vessels should not have been a source of external beta exposure.

Krypton-85, a beta and gamma emitter, was used in two leak detection units as part of the Pinellas 

Quality Control Program. The leak detection units 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 is a whole-body emitter with the possibility of a skin dose from 

beta radiation.

Small sealed sources of plutonium produced by Los Alamos National Laboratories were shipped to 

the Pinellas Plant for use in the RTGs. Each sealed source contained approximately 80% 238Pu and 

20% 239Pu. The plutonium sources were enclosed in a triple metal encapsulation. These units were 

never opened. The capsules were inserted as heat sources for the RTGs (DOE 1983, pp. 2-11). The 

first recorded receipt of plutonium was on January 18, 1957, when the Plant received a 7-g (4.44 × 

1012-Bq or 120-Ci) 238Pu source to calibrate Health Physics monitoring equipment. Seven plutonium 

heat sources, totaling 54.4 g (3.515 × 1012 Bq or 950 Ci) of 238Pu, arrived at Pinellas from Sandia 

National Laboratories on November 4, 1975. About 50 RTG generators were produced every month 

(Burkhart 1987a, p4). The last on-site plutonium, with the exception of calibration check sources, was 

removed from the Plant in February 1991 (GEND 1993).
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Sources of ionizing radiation at the Pinellas Plant included other low-activity radioactive sources, such 

as those used to check or calibrate radiation detectors (i.e., calibration sources) and analytical 

devices employing radioactive byproduct material or X-rays produced by a radiation generating device 

(RGD). The calibration sources were maintained in the Health Physics Laboratory in Area 113 of 

Building 100. While the exact inventory of radioactive sources varied over time, most were sealed 

microcurie check sources of radioactive isotopes such as 137Cs, 238Pu, 239Pu, 60Co, 14C, and 90Sr. 

These low-activity radioactive sources could have included alpha, beta, photon, and neutron emitters. 

For example, small quantities of 14C-labeled solvents were used in a laboratory testing operation. The 

largest 14C source was a 2.59 × 107

-Bq (700-μCi) source used for liquid scintillation counting 

calibration. Radiation doses from 14C result from internal deposition (DOE 1983). The analytical 

devices could have used radioactive byproduct material such as 85Kr, 109Cd, or 241Am that requires 

U.S. Nuclear Regulatory Commission certification. Analytical devices with an RGD would have 

appropriate radiation shielding to comply with performance regulations of the U.S. Food and Drug 

Administration as specified in 21 C.F.R. §§ 1020.30, 1020.31, and 1020.40 (FDA 2004) along with 

Occupational Safety and Health Administration occupational exposure regulation 29 C.F.R. § 

1910.1096 (DOL 2003). (Note: The current regulations might have changed in surface radiation levels 

since 1957.) However, there were recorded incidents at Pinellas involving these types of devices. A 

listing of unusual events and incidents is included in Table 2-4 of the facilities and processes section.

Table 6-1 lists the X-ray generating equipment and its locations on the Pinellas Plant site. All 

analytical devices used at the Plant were of the types and source strengths typically used by 

mainstream industrial or process-related users. 

Table 6-1. X-ray producing equipment. 

Location Quantity Type ID Number

107 2 X-ray emission (XRE) units CCN87103, CCN99591

114 2 Industrial X-ray units MN87904, MN87851

114 1 Faxitron X-Ray Corporation MN87243

114 1 TFI X-ray television None found

127 1 Inspector unit None found

138 1 Electron beam welder MN62183

161B 2 X-ray diffraction (XRD) units MN86640, MN86641

161B 1 Micro XRD unit MN86747

161B 1 Electron microprobe MN86607

161B 1 Energy dispersive analyzer MN86475A

162 2 Scanning Electron Microscope (SEM) MN94109, MN94225

162 1 Transmission Electron Microscope MN94168

163 1 X-ray thickness gauge MN550042

164 1 Cabinet X-ray unit CCN91407

175 1 XRD unit CN87952

176 1 XRE unit None found

192B 1 Sedigraph MN94285

193N 1 SEM MN91992

194E 1 Picker (cabinet X-ray unit) MN87042

300 2 Electron beam welder MN61660, MN76803

300 1 Phillips (cabinet X-ray unit) MN87810

300 2 Faxitron X-Ray Corporation MN088001, MN099915

400 2 Electron beam welder MN61346, None found

400 1 Cabinet X-ray unit MN63294

Warehouse 1 X-Ray television None found

Source: GEND 1987a, GEND 1986

An ion accelerator facility that was used to test various components generated photons (up to 200 

KeV) and neutron (likely 2.5 or 14 MeV) radiation (located in building 800). Commercial neutron 
Document No. ORAUT-TKBS-0029-6 Revision No. 00 PC-1 Effective Date: 08/03/2006 Page 10 of 42

generators producing neutron energies of 2.5 or 14 MeV neutrons were also used to calibrate or 

perform QA checks of various components mostly located in building 100.

6.1.4

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

working in neutron generator production areas. Table 6-2 lists the total number of Pinellas employees 

and the number of employees monitored for radiation exposure for years with available data. 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 dosimetry (377 of an average yearly labor force of 

1,372). During the 1980s, while the data are not completely available, from 370 to approximately 400 

of 1,650 to 1,975 workers (approximately 20%) were monitored for radiation dose. No documentation 

was found to show that all employees were monitored at some time during Pinellas operations. For 

the majority of Pinellas operations, external dosimetry was exchanged and analyzed monthly. 

Beginning in January 1990, external dosimetry was exchanged and analyzed quarterly (Burkhart 

1988, GEND 1990a). It is evident that not all groups of workers were required to wear dosimeters 

(Public outreach 2004). As part of the EEOICPA dose reconstruction, all recorded doses for claimants 

are received from Pinellas for analysis. 

Occupational Dosimetry Program Overview

Table 6-2. Personnel radiation dosimetry from AEC annual reports.

a

Year

Number of PAO-AEC (or DOE) 

and GEND employees

Number of monitored 

employees

Number of employees in a given dose range

< 1 rem 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 1,397 281 281 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 NAb NA NA NA

1975 NA NA NA NA

1976 NA 317 317 0

1977 NA 300 300 0

1978 NA 297 297 0

1979 NA 334 334 0

1980 NA 376 376 0

1981 NA 389 389 0

1982 NA 408 408 0

a. Source: Data from Form AEC-190.

b. NA – not available.

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 from dosimeters was recorded at the time of measurement, reviewed by 

Pinellas health physicists, and routinely made available to workers. External Dose Reconstruction 

Implementation Guidelines (NIOSH 2002) indicates that these records represent the highest quality 
Document No. ORAUT-TKBS-0029-6 Revision No. 00 Effective Date: 09/15/2005 Page 11 of 42

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

the available records.

A basis of comparison for dose reconstruction is the Personal Dose Equivalent, Hp(d), where d 

identifies the depth (in millimeters) and represents the point of reference for dose in tissue. For weakly 

penetrating radiation of significance to skin dose, d = 0.07 mm and is noted as Hp(0.07). For 

penetrating radiation of significance to whole-body dose, d = 10 mm and is noted as Hp(10). 

6.2 DOSE RECONSTRUCTION PARAMETERS

6.2.1

Between 1957 and 1974, GEND-Health Physics conducted radiation dosimetry management and 

analysis through in-plant processing of X-ray and neutron-sensitive photographic films (Burkhart 

1987b). In 1974, GEND contracted with R. S. Landauer Jr. & Co. (Landauer) to be the principal 

supplier and processor of the dosimetry that GEND used. This arrangement continued until Pinellas 

Plant operations ended.

Historical Administrative Practices

The available information indicates that, from the start of nuclear operations in 1957, the Pinellas 

Plant used film badge dosimetry available from or similar to the type sold by Nuclear-Chicago for the 

designated radiation control areas (GEND 2004a). Beginning in 1974, the Plant began using 

dosimetry provided exclusively by Landauer. The original Landauer dosimetry was based on film 

badge technology (Ward 1974). 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). By the 1980s, GEND 

was using Landauer Model E1, G1, G5, and U3 badge dosimetry.

From October 1979 to October 1987, the Plant used albedo dosimeters from Mound Laboratories for 

evaluating exposures to 238PuO2 2-MeV average neutrons and gammas during handling of RTG units 

at 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 U.S. Department of 

Energy’s 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 dosimetry program at the Pinellas Plant.

There is no consistent documentation on how GEND processed dosimetry for the first 18 years of 

operation. Individual accounts indicate that the Pinellas Plant could have processed dosimetry during 

this time. Beginning in 1974, Landauer processed the dosimetry and provided exposure reports to 

GEND for review after badge processing (Figure 6-1 is a replication of portion of a report (Landauer 

1987). Information in the Landauer reports included personnel data (identification number, name, and 

social security number), 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 Pinellas). The exposure information was entered in 

each person’s exposure history by hand or in a Pinellas-based computer system and the General 

Electric Health and Safety Record (HSR) system on ionizing radiation (Richards 1986). Attachment A 

contains examples of the handwritten and computer records for 1960 to 1988. The forms used in 1960 

were also used before 1960 and the forms used in 1988 were used until Pinellas ceased operations.
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Table 6-3. Pinellas Plant historical dosimetry events. 

Date Event Reference

April 1957 New employee orientation in radiation safety offered. GEND 1990a

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

November 1957 Measured neutron doses of neutron generators at 10 mrem/pulse 

at 1 in.

GEND 1990a

December 1957 Sandia National Laboratories-Albuquerque asked to provide film 

badges.

GEND 1990a

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

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

number of employees.

GEND 1990a

February 1965 Memorandum comparing performance of two types of neutron 

badges and two types of gamma badges.

Szedziewski 1965

Late 1969 Film badge fading study. GEND 1969

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

Third Quarter 1974 Began using Landauer for source of film badges. GEND 1990a

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.

October 1986 Memorandum on estimated doses to GEND personnel handling 

unmarked neutron generator units.

Burkhart 1986

October 1987 End use of Mound neutron dosimeters.

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

April 1990 Changed from Landauer film badges to TLD dosimetry.

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

generators.

GEND 1996

September 1994 Nuclear product development and testing end. Other Radiation 

related work did continue. 

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 (GENDa 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. Analysis of the available 

claimant records found no documentation that this reporting requirement was ever exceeded during 

the time Landauer provided dosimetry services to the Plant (1974 to 1994).
Document No. ORAUT-TKBS 0029-6 Revision No. 00 Effective Date: 09/15/2005 Page 13 of 42

Figure 6-1. Regenerated example of a Landauer dosimetry report from the 1970s. 

GENERAL ELECTRIC COMPANY ACCOUNT NO. 70463

NEUTRON DEVICES DEPT

ENVIRONMENTAL HLTH

7887 BRYAN DAIRY RD

LARGO FL 33543

Participant 

ID 

number Name

Social 

security 

number

Note 

(see 

reverse 

side)

Dosimeter 

type Use

Radiation 

quality

Exposure to 

badge (millirems) 

for period(s) 

indicated below

Cumulative totals 

(millirems)

Adjustments Sex

Birth date

Number 

badge reports

Inception 

date

Calendar 

quarter

Year to 

date Permanent

Deep Shallow Deep Shallow Deep Shallow Deep Shallow MO DA YR To date Qtr MO YR

[PIR] AREA MONITOR 183 E 1 M M M M M M M M 11 1 11

[PIR] [PIR] [PIR] U 3 M M M M 24 2 11 84

[PIR] CONTROL G 1 M M M M M M 260 260 295 5 6 74

[PIR] TLD CONTROL M M M M 101 3 2 79

[PIR] LOAN U 3 M M M 220 96 3 2 79

[PIR] [PIR] [PIR] G 1 M M M M M M M M 7 5 9 86

[PIR] [PIR] [PIR] G 1 M M M M M M 100 100 M 256 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M M M 250 250 M 256 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M M M 110 110 F 256 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M M M 10 10 23 3 11 84

[PIR] [PIR] [PIR] G 5 M M M M 23 3 2 85

[PIR] [PIR] [PIR] G 1 M M M M M M 240 240 M 256 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M 30 30 250 250 M 255 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M 40 40 90 90 M 256 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M M M 90 90 M 256 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M M M 200 100 M 256 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M M M 170 170 M 257 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M M M 180 180 M 258 5 6 74

[PIR] [PIR] [PIR] G 1 M M M M 40 40 140 140 M 254 5 6 74

PIR = privacy information removed.
Document No. ORAUT-TKBS-0029-6 Revision No. 00 Effective Date: 09/15/2005 Page 14 of 42

6.2.1.1 Performance Testing

Neutron Studies

Eastman Kodak nuclear emulsion type A (NTA) film was used for neutron measurements and should 

be similar to the Landauer NTA dosimetry used at Pinellas after 1974 for the Type J badge (GEND 

1969, Koperski 2004). As stated in the Hanford, INEEL, and Nevada Test Site Occupational External 

Dose TBDs, NTA was basically the only common dosimeter method available to measure neutron 

dose in AEC facilities prior to 1978. Results reported at the first AEC Neutron Dosimetry Workshop in 

1969 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 et al. 1969). The neutron spectra at Pinellas were known to be dominated by 14-MeV 

deuterium-tritium fusion neutrons due to the unique design of the neutron generators. Lower energy 

fusion neutrons from deuterium-deuterium reactions of 2.5 MeV were also likely to occur. In addition, 

the method(s) used to calibrate the NTA film (Landauer or other dosimetry supplier) is not known. 

Typically, the dominant neutron exposure from nuclear weapon components at GEND was readily and 

reliably measured with NTA film dosimeters. For neutrons from RTG production, Pinellas initially 

continued with NTA film but changed to polycarbonate dosimeters in 1978; Mound albedo dosimeters 

were used in the 1980s.

Specific Pinellas Plant Studies

The NTA film badge neutron dosimeters used from 1957 through 1977 underwent track fading during 

use, causing a loss of information (Holliday 1978). At least for a period, the neutron dosimeters were 

made of Kodak Type A film (GEND 1969). GEND established a factor of 3 to correct for track fading 

beginning in January 1970, based on a track fading study (GEND 1969). 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/week; 

while the dose assigned to a worker receiving a total exposure on the last day of the monitoring period 

would be 40% of the true dose for a monthly wear period (GEND 1969). The Landauer polycarbonate 

badge that replaced these badges in 1978 did not undergo track fading, so no correction factor was 

applied after 1978 (Holliday 1978). The Landauer I8 Neutrak TLD was limited with sensitivity to RTG 

neutrons, calculated to be 67% of the true dose by GEND health physics personnel.

6.2.1.2 Reports

An as-low-as-reasonably-achievable (ALARA) Program report discussing occupational exposures was 

submitted to the manager of the Environmental Health and Safety Program. Attachment A contains 

examples of the reports kept by the Environmental Health and Safety Program and its radiation 

protection program from 1960 to 1988 for one claimant.

6.2.2

This analysis was unable to locate specific designs of the film dosimeters used for approximately the 

first 20 years (1957 to 1974) at the Pinellas Plant, and there is limited documentation that indicates 

there was an early relationship with Nuclear-Chicago (GEND 2004a). However, the dosimetry type 

and sources are well documented in the Landauer exposure reports from 1974 until the Plant ceased 

operations. Table 6-4 summarizes the monitoring technique and describes expected and known Plant 

dosimeters. Table 6-5 summarizes the Minimum Detection Level (MDL) of these dosimeters along 

with the maximum potential missed dose (NIOSH 2002). The MDL, which is widely used in other 

documents prepared for the NIOSH Dose Reconstruction Project, can vary depending on dosimeter 

type, processing equipment, calibration techniques, and procedures. Because of these variations, a 

Dosimetry Technology
Document No. ORAUT-TKBS-0029-6 Revision No. 00 PC-1 Effective Date: 08/03/2006 Page 15 of 42

review of typical MDLs for photon dosimeters was conducted and is documented in ORAU 2004a. 

The typical maximum missed dose per exchange cycle for photon dose for film dosimeters is 40 mrem 

for DOE facilities in general.

6.2.3

GEND 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) Personal Dosimetry Performance – Criteria for 

Performance (ANSI 1993). Each test used approximately six to nine badges.

Dosimetry Calibration Practices

For the G1 film emulsion package, badges were exposed to the Shepherd model 81-12 137Cs beam 

irradiator in Building 800 (a gamma check for 662-KeV photons). For beta calibration, the badges 

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 badges to keep them free from uranium 

contamination.

Calibration of the E1 polycarbonate badges 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. 

The R1 film badge, TLD albedo badge, and CR-39 badge were placed on a water phantom and 

exposed to a Shepherd model 149 241Am/Be calibrator. 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.

Further details of the dosimetry are listed in Table 6-5. This includes the dosimeter types along with 

some dosimeter configuration and energy response characteristics. Table 6-5 lists the associated 

MDLs and the maximum potential missed photon or beta dose. Even though the dosimeter 

designations or types changed, the MDLs did not vary much over the operational history of the 

Pinellas Plant. 

However, even though there is GEND 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 were 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 from the personnel packets, the densities on the personnel film were 

normalized to Landauer controls only and a non-minimal 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 (Koperski 2004).
Document No. ORAUT-TKBS-0029-6 Revision No. 00 PC-1 Effective Date: 08/03/2006 Page 16 of 42

Table 6-4. Dosimetry used at Pinellas Plant for external whole-body, wrist and extremity exposures. Period Monitoring technique Dosimeter description

Beta/photon Dosimeters

1957 – June 197

whole body

Photographic film badge. Nuclear-Chicago or similar film badges utilized in-house. Nuclear-Chicago film 

badge contained single film packet. Three filters (front and back) were incorporated 

into film badge for energy dependence: cadmium, aluminum, and lead (ORAU 

2003b).

July 197 – About 1983*

whole body

Landauer Type K (neutron, 

beta-gamma), J (beta and 

gamma).

Type J and K dosimetry were film badges but the K dosimeter had NTA film for 

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

MeV.

About 1974* – April 1990

whole body

Landauer G1. Film emulsion packaged placed in standard Gardray holder/badge for monitoring 

beta, X-ray, and gamma exposure. Insensitive to neutron radiation. Required in 

areas where krypton-85 was used. Required for radiation-generating equipment 

and accelerator operators.

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

May 1990 – 1997

whole body

Landauer Z1 dosimeter. Comprised of 3 TLD-700 chips for monitoring beta, X-ray and gamma exposure. 

Insensitive to neutron radiation. Replaced Landauer G1.

July 197 – About 1983*

Wrist

Landauer Type M (wrist 

beta-gamma) badges.

Type M dosimetry was a film badge. Gamma and X-ray: 30 keV to 20 MeV; beta: 

over 1.5 MeV.

About 1983* – 1990

Wrist

G5 wrist film badge. Responded to beta, X-ray, and gamma exposure to provide data on extremity dose. 

Extremity dosimetry was worn in locations where plutonium oxide was handled.

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

1991 – 1997

wrist

K5 TLD wrist badge. Comprised of 3 TLD-100 chips.

1957-1974 Film badge-finger. Nuclear-Chicago or similar film badges. Nuclear-Chicago film badge contained 

single film packet. Three filters (front and back) were incorporated into film badge 

for energy dependence: cadmium, aluminum, and lead (ORAU 2003b).

About 1983* – 1997

finger ring

U3 TLD (LiF). Responded to beta, X-ray, and gamma exposure to provide data on extremity dose. 

Extremity dosimetry was worn in locations where plutonium oxide was handled.

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

Neutrons

1957– 1978

whole body

NTA film badge. Nuclear-Chicago or similar film badges Kodak Nuclear Track Emulsion NTA films: 

Fast neutrons undergoing elastic collision with content of emulsion or cellulose 

acetate base material produce recoil protons, which are recorded as photographic 

tracks in emulsion. Track density is linear function of dose. Developed image 

exhibits tracks caused by neutrons, which can be viewed using appropriate imaging 

method (i.e., oil immersion) and 1000X power microscope or projection capability.

July 1974 – About 1978*

whole body

Landauer Type K badge. Landauer NTA film badge for neutrons 1 to 10 MeV.

About 1978* – 1997

whole body

Landauer Neutrak E1 

dosimeter. 

Polycarbonate (lexan) neutron recoil track registration device used to monitor fast 

neutron interactions. Neutrak 144 has 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. E1 could be combined with G1 (and later with Z1). 

Accelerator operators were required to wear E1/G1 combination. Workers were 

required to wear E1/G1 dosimeters when working around neutron generators. 

E1/G1 dosimeter or G1 dosimeter was required when working with calibration 

sources. 

October 1979 – October 

1987

whole body

Mound albedo dosimeter. Evaluated exposures to 238PuO2 2-MeV average neutrons and gammas during 

handling of RTG units at Building 400.

October 1987 – 1994

whole body

Landauer I8, I1 or RI 

Neutrak Extended Range 

dosimeter. 

Combined TLD albedo neutron monitor with 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 above-described elements. Qualitative relationship was derived to determine 

ratios of neutrons of various energies. This badge was combined with G1 to make 

dosimeter, known as R1 that monitored beta, X-ray, gamma, and neutrons. R1 film 

badge arrangement, which was used in locations where 238PuO2 was handled, 

replaced Mound albedo dosimeter. I8 did not meet all DOELAP requirements during 

performance testing. After G1 was replaced by Z1 (Z1 was called F1 beginning in 

1995), combined unit was known as I1. After production of RTGs was halted in 

October 1991, I1 was only used as area monitor for americium-beryllium (AmBe) 

source. I8 is still offered by Landauer. 

Sources: Burkhart 1987b, 1988; GEND 1990a,b; Greene 1985; Holliday 1978; Landauer 2004; Ingle 1991; Weaver 1991, 1995, 1996; 

Passmore 2004; Koperski 2004.
Document No. ORAUT-TKBS-0029-6 Revision No. 00 PC-1 Effective Date: 08/03/2006 Page 17 of 42

* This analysis found no documentation that shows the start of G1, U3, G5, I8, and E1 dosimeter use and the end of Type K, J, and M 

badges use.

Table 6-5. Minimum Detectable Level (MDL) and Maximum Potential Missed Photon or Beta Dose. 

Dosimeter Period of Use MDL (rem)a Max Annual Missed Dose (rem)

(MDL/2)

Nuclear Chicago or Pinellas 

film badge–whole body

1957- June 1974 0.04 photons

0.04 beta

0.24 beta -photons (monthly)

Landauer Type K 

whole body

July 1974-About 1983b 0.01 photons 

0.04 beta

0.12 photons (monthly)

0.24 beta (monthly)

Landauer Type J

whole body

July 1974-About 1983b 0.01 photons

0.04 beta

0.12 photons (monthly)

0.24 beta (monthly)

Landauer G1- 

whole body

About 1983b – April 1990 0.01 photons 

0.04 beta

0.06 photons (monthly)

0.02 photons (quarterly)

0.24 beta (monthly)

0.08 beta (quarterly)

Landauer Z1 dosimeter - 

whole body

May 1990- 1997 0.01 photons

0.04 beta

0.06 photons (monthly)

0.24 beta (monthly)

Nuclear Chicago or Pinellas 

film badge Film-badge-wrist

1957- June 1974 0.04 photons

0.04 beta

0.24 beta-photons (monthly)

Landauer Type M - 

Wrist

July 1974-About 1983b 0.01 photons 

0.04 beta

0.06 photons (monthly)

0.24 beta (monthly)

G5 wrist film badge – 

Wrist

About 1983b – 1990 0.01 photons

0.04 beta

0.06 photons (monthly)

0.04 photons (quarterly)

0.24 beta (monthly)

0.08 beta (quarterly)

K5 TLD 

wrist 

1991-1997 0.01 photons

0.04 beta

0.06 photons (monthly)

0.24 beta (monthly)

Nuclear Chicago or Pinellas 

film badge Film badgefinger

1957- June 1974 0.04 photons

0.04 beta

0.24 beta-photons (monthly)

U3 TLD (LiF) - 

finger ring

About 1983b – 1997 0.03 photons

0.04 beta

0.18 photons (monthly)

0.08 photons (quarterly)

0.24 beta (monthly)

0.08 beta (quarterly) a. Estimated MDLs for commonly used photon dosimetry (ORAU 2004a). b. Dates are approximate. No information could be found regarding when the G1 replaced the Type K and J, when the G5 replaced the Type 

M, or when use of the U3 began. 

6.2.4

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

radionuclides and machine-generated X-rays, neutrons, and electrons. Radionuclides were used 

directly in the manufacturing processes, manufacturing support, and various calibration sources. 

Machine-generated X-rays from the equipment listed in Table 6-1 supported the Quality Assurance 

program and process monitoring, and provided specific manufacturing support (i.e., E-beam welding 

and operation of an accelerator).

Workplace Radiation Fields

Table 6-6 lists radionuclides that would contribute to workplace radiation fields. The radionuclides 

applied in the manufacturing processes at the Pinellas Plant were kept in containers, in sealed 

sources, or in the process piping. 
Document No. ORAUT-TKBS-0029-6 Revision No. 00 Effective Date: 09/15/2005 Page 18 of 42

Because most of the primary radiation from these sources (alphas and betas) would not penetrate the 

containers, sealed source encapsulation, or the process piping, and the workers would not be in close 

contact with the sources, the expected radiation fields would be very small. In the cases of tritium and 

85Kr, which were gases in the Pinellas processes, vent hoods or direct connections to ventilation 

exhaust systems quickly removed the gases from worker spaces. The largest calibration source was a 

120-Ci 137Cs source used in a Model 81-12 Beam Calibrator manufactured by J. L. Shepard and 

Associates (GEND 1977). Three neutron sources were used at various times, two of which contained 

239Pu (1.7 Ci and 0.43 Ci) and an AmBe source with 10 Ci of 241Am. All three were removed from the 

Pinellas site by the end of 1991 (1979, 1990, and late 1991, respectively). All other calibration sources 

were in the mCi and μCi ranges and were sealed, and would not contribute to occupational doses 

(other than to workers who used the mCi sources) (GEND 2004b).

Table 6-6. Radionuclides contributing to workplace radiation fields. 

Nuclide Source Half-life

Energies and abundances of major radiationsa

Alpha (MeV) Beta (keV) Gamma (MeV) Neutron (MeV)

H-3 Loaded in neutron 

generators

12.33 yr --- 18.6 X-rays were also 

produced 

probably of 

medium energy – 

30 –250 kev.

14 MeV and 2.5 MeV neutrons from 

testing of D-T neutron generators

2.5 MeV neutrons from D-D and D-T 

neutron generators.

Kr-85 Leak check source 10.72 yr --- 687.1 (99.6%) 0.514 (0.4%) ---

Cs-137 Calibration source 30.07 yr --- 1175.6 – (5.4%)

511.6 (94.6%)

0.6617 (decay of 

Ba-137m)

---

Am-241 Part of AmBe sealed 

neutron source

432.2 yr 5.49 (85%)

5.44 (12.8%)

5.39 (1.4%)

5.54 (0.3%)

--- 59 (35.9%)

13.9 (42.7%)

AmBe produces thermal neutrons by α-n 

reaction. 

Depleted uranium metal powder for H-3 storage

U-238 99.75% weight 4.51E9 yr 4.15 (21%)

4.20 (79%)

--- --- Weak spontaneous fission (2 MeV).

U-235 0.25% weight 7.1E8 yr 4.21 (6%)

4.37 (17%)

4.40 (55%)

4.60 (5%)

--- 0.144 (11%)

0.163 (5%)

0.186 (57%)

0.205 (5%)

Weak spontaneous fission (2 MeV).

U-234 0.0005% weight 2.47e5 yr 4.72 (28%)

4.77 (72%)

--- 0.053 (0.12%) ---

Plutonium isotopes used in RTG enclosed in triple metal-encapsulated sealed sources

Pu-238 80% 87.74 yr 5.50 (72%)

5.46 (12%)

--- --- Weak spontaneous fission (2 MeV Avg)

α, 18O reaction.

Pu-239 20% 24,110 yr 5.16 (88%)

5.11 (11%)

--- --- Spontaneous fission (2 MeV)

α, 18O reaction. 

a. Energy information from Hacker 2001.

The only open-area radiation fields to be encountered by workers would be from the testing of neutron 

generators and the use of machinery-generated X-rays. This was caused by locating a neutron 

generator test stand or X-ray generating equipment in a room. 

For neutron generator tests, there was no shielding around the test area with the exception of the 

room walls. X-ray generating equipment was designed to be in a shielded enclosure that, by 

certification, was below 0.5 mR/hr at 5 cm from the surface of the device. Most devices would have a 

lower radiation field. Shielding around the test area was added later at an unspecified time. There was 

also X-ray production from the tritium tube testing. This is implied by a description of exposures in a 

GE memo (Greene 1984) and in a memo from Sandia lab describing X-ray generation from neutron 

tube testing. By calculation, it is assumed that X-ray dose was linear to the beam current of electrons 

and ions. The dose equivalent of X-rays produced was close to that of the neutron dose equivalent 

without any shielding of the tube present (Brainard 1991).
Document No. ORAUT-TKBS-0029-6 Revision No. 00 PC-1 Effective Date: 08/03/2006 Page 19 of 42

To support the production of the RTGs, Pinellas health physicists measured RTG radiation fields to 

yield 3 mrem/hr gamma and 10 mrem/hr neutron at 10 cm (approximately 4 in.) from the surface of 

the device and measurements during a 12/17/1987 hazard review yielded 20 mrem/hour gamma and 

60 mrem/hr neutron at contact. Both measurements were most likely from one RTG unit (GEND 

1987c). Another indicator of the expected radiation fields experienced by Pinellas workers comes from 

a review of the dosimetry records in the NIOSH Office of Compensation Analysis and Support (OCAS) 

Claims Tracking System (NOCTS) database of the 285 claimants for external exposures. Of the 285 

claimants, 60 have Pinellas employment with whole-body external doses from all forms of radiation 

greater than 1 mrem, with a total of 18.140 person-rem. Of this total dose, 3.085 rem (17%) are from 

intake of tritium, 7.152 rem (39.4%) are from neutron exposure, 6.587 rem (36.3%) are from 

beta/gamma exposure, and 1.317 rem (7.3%) are for extremity doses from beta/gamma exposure. Of 

the 60 claimants with external exposures, 34 have lifetime doses greater than 100 mrem, and nine 

have Pinellas lifetime doses from all sources greater than 500 mrem. Most of these doses were 

accumulated in a sporadic manner by receiving a dose over a fraction of a year and/or not having any 

recordable doses for several months or for years in some cases. Because a limited number of 

monitored employees received extremity dosimetry, only 5 of the 60 claimants with external doses 

have recorded lifetime extremity doses (20, 28, 56, 483, and 730 mrem). 

The claimants’ exposure histories are supported by the annual radiation exposure summary reports 

compiled by the AEC from 1960 to 1973, which describe two cases in which a Pinellas worker 

received more than 1 rem for the year.

6.2.4.1 Beta-Gamma Exposures

The NIOSH Interactive RadioEpidemiological Program (NIOSH 2002) discusses three photon energy 

bands – below 30 keV, 30 to 250 keV, and above 250 keV. The principal source of photons to the 

greater number of Pinellas workers would have been the various X-ray machines due to the limited 

access by a small number of workers to direct contact with radionuclides that emit photons (85Kr in the 

two Radiflo leak detectors, the one 137Cs sealed source in a shielded calibration machine, depleted 

uranium in the tritium storage cylinder, and 238PuO2 RTG heat sources). 

The X-ray machines would generate photon energies below 250 keV. Because the machines were 

shielded, those used probably have a range of energy settings; and because measurements of the 

photon energies on the outside of the shielding are not available, this analysis grouped this source of 

photons in the 30- to 250-keV energy interval, which should be claimant-favorable. Table 6-7 lists beta 

and photon energies and percentages for processes at the Pinellas Plant.

Table 6-7. Beta and photon radiation energies and percentages for Pinellas Plant processes. 

Facility or location Process type

Radiation 

type

Energy 

interval Percentage

Buildings 100, 200, 

300, 400, Medical

Neutron generator production, 

quality assurance, and RTG 

production.

Photon 30 - 250 keV 100

Building 100, Area 109 Product analysis (Radiflo and Xray). Beta

Photon

> 250 keV

30 – 250 keV

> 250 keV

100

86

14

Building 800 Calibration and accelerator. Photon 30 - 250 keV 100

External exposure due to beta particles is unlikely. The predominant source of beta at the Pinellas 

Plant was tritium, but this radionuclide emits a low-energy beta of less than 30 keV. Such a low- 
Document No. ORAUT-TKBS-0029-6 Revision No. 00 Effective Date: 09/15/2005 Page 20 of 42

energy beta would not penetrate the dead layer of the skin. The only sources of beta particles with 

sufficient energy to penetrate the skin are 85Kr and 137Cs (Note, there was the use of depleted uranium 

(DU) beds for tritium processing but the DU was contained in small metal cylinders that would have 

attenuated the beta component from the worker). Several incidents occurred in which workers were 

exposed to 85Kr, so beta should be considered in the area where the two Radiflo units were located. 

The 137Cs was in a sealed source in a shielded cabinet in the concrete Building 800, so the probability 

of worker beta exposures was remote. 

Beta exposures were possible for X-ray diffraction and electron beam devices if containment of the 

beams were compromised. It is more likely that exposures would have been from X-rays or 

bremsstrahlung production and not from any free electron beam. The exposures, if diffuse, would 

have been monitored by film badge or TLD monitored personnel. Most unusual occurrences were 

documented and any likely penetrating or non-penetrating exposures would have been addressed in 

personnel medical records. 

6.2.4.2 Neutron Exposures

There were two distinct neutron sources at the Pinellas Plant: 2.5 and 14-MeV neutrons from the 

testing of neutron generators and 2-MeV average neutrons from the plutonium sealed sources in the 

RTGs. The locations of the operations producing neutrons include a small number of areas. However, 

personnel wore neutron dosimetry in other locations; Table 6-8 lists the locations where workers wore 

neutron dosimetry and the expected neutron source energy.

Table 6-8. Selection of neutron radiation energies for 

Pinellas Plant processes. 

Facility or 

location Process type

Neutron 

energy

107 Tube assembly 2-14 MeV

128 Tube test 2-14 MeV

131 Final test 2-14 MeV

132M Fan room 2-14 MeV

182 Tube assembly 2-14 MeV

183 General development 2-14 MeV

184 Tube testing 2-14 MeV

191 CPE hood room 2-14 MeV

194 Engineering environmental testing 2-14 MeV

Radioanalytical Laboratory 2-14 MeV

Building 200 Product QA testing 2-14 MeV

Building 300 2-14 MeV

Building 400 RTG assembly and testing 0-12 MeV

Building 800 Calibration and accelerator 2-14 MeV

Medical 2-14 MeV

The RTG PuO2 heat source spectrum is illustrated in figure 6-2. The spectrum was probably obtained 

from the mound plant and was used to analyze the effectiveness of various Landauer’s Neutrak TLND 

dosimeters by the Pinellas health physics department (Burkhart, 1987a). It was determined that the 

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

source spectrum. 
Document No. ORAUT-TKBS-0029-6 Revision No. 00 Effective Date: 09/15/2005 Page 21 of 42

Figure 6-2.a RTG Plutonium Source Neutron Energy Spectrum – PuO2

Microspheres. a. (Burkhart, 1987a) 

6.3 ADJUSTMENTS TO RECORDED DOSE

Adjustments to Pinellas Plant recorded doses are necessary to arrive at a claimant-favorable dose, 

considering the uncertainty associated primarily with the complex workplace radiation fields and 

exposure geometries. A key item for dose reconstructors to understand about the GEND radiation 

protection program is that for an individual lifetime dose (denoted as the “NDD” lifetime dose in GEND 

files), the internal exposure from tritium was combined with the external exposure from neutrons, 

gammas, and X-rays. To reconstruct a claimant’s lifetime dose properly, the dose reconstructor 

should separate the tritium dose from the rest of the external dose. As discussed in Section 6.2.4, 60 

claimants have recorded lifetime doses equal to or greater than 1 mrem. Table 6-9 lists the results of 

a review of their records that indicated external exposure determined the segregation between doses 

from tritium, neutrons, photons, and extremity.
Document No. ORAUT-TKBS-0029-6 Revision No. 00 Effective Date: 09/15/2005 Page 22 of 42

Table 6-9. Internal and external doses among 

Pinellas claimants (mrem). Total Pinellas 

lifetime dose

Tritium 

dose

Neutron 

dose

Photon 

dose

Extremity dose, 

if available

3,770 87 218 3,465

1,111 1,111 0 0

1,054 705 271 78

914 15 632 267 28

802 33 60 709

791 2 601 178 483

599 236 283 80

519 379 140 0

506 1 244 256

494 174 214 106

480 0 110 370 730

461 15 347.5 98.5

357 0 357 0

350 4 346 0

320 8 216 87

317 31 247 39

133 0 133 0 

264 2 123 139 56

262 0 262 0

248 0 218 30

241 0 231 10

227 1 201 25

220 0 220 0

219 16 100 103

209 0 82 127

191 19 140 32

189 11 161 17

183 0 183 0

172 14 126 32

140 140 0 0 

127 0 74 53

120 0 120 0

100 0 100 0

83 3 0 80

74 0 14 60 20

65 0 0 65

61 1 60 0 

52 8 44 0

37 23 14 0

47 N/A N/A N/A

46 0 46 0

40 0 40 0

35 0 35 0

35 0 35 0

30 0 0 30

30 0 0 30

27 0 27 0 

22 2 20 0

21 5 0 16

20 0 20 0 

17 0 17 0

14 0 14 0

11 9 0 2

9 9 0 0

8 8 0 0

8 8 0 0

5 0 5 0

4 4 0 0

2 0 0 2

1 1 0 0 

N/A – not applicable. Highest case number used in this table was case #16921.
Document No. ORAUT-TKBS-0029-6 Revision No. 00 PC-1 Effective Date: 08/03/2006 Page 23 of 42

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

neutron and photon doses. 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 relative 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 upon neutron and photon exposure rates from the processing of the RTG PuO2 heat sources

as measured in 1987 (Forest 1974). Exposure rates measured at another time were 0.37 mR/hr 

neutron and 0.18 mR/hr gamma at 24 inches. Actual dose equivalent received form removal and 

processing of 15 units for one worker (apparently the typical workload was 50 generators / 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 (GEND 1983). Since the RTG radioisotopic ratios and 

activity remained unchanged and the work- stations and processes consistent, this ratio can also be 

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

6.3.1

No adjustments to photon doses (including X-rays) are known to have been made for Pinellas 

dosimetry. No adjustments to photon doses are necessary for dose reconstruction with the exception 

of those specified as part of defining the missed dose.

Photon Dose Adjustments

6.3.2

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 – 14 MeV (ICRP 1990). This value 

is higher than other values that could have been used (such as those in NCRP 1971).

Neutron Dose Adjustments

The plutonium in the RTGs emits neutrons with an energy range of thermal to 12 MeV (Figure 6-2) 

and a 2 MeV average. The available GEND documentation does not describe any separate 

processing of the dosimetry applied for RTG manufacture that is different from the other neutron 

dosimetry. Therefore, this analysis assumed that the RBE of 10.0 was applied for the RTG neutron 

source, which would be consistent with ICRP (1990).

The ICRP 60 weighting factor is 10 for the neutron energy band from 2 to 14 MeV for neutron 

generators, which is equal to the value that Pinellas used historically. The highest ICRP 60 weighting 

factor is 20 for the neutron energy band from 0 to 2 MeV, which comprises about 25% of the neutrons. 

Table 6-10 represents the neutron energy bins and associated dose fractions to be used for the RTG 

PuO2 heat source workers. The most reasonable correction factor for converting the ICRP 38 to the 

ICRP 60 neutron radiological weighting factor would be 1.47. 

A correction for the under-response of NTA film to lower energy neutrons of < 500 keV including 

albedo neutrons is based on 67% track fading per month for NTA film, 33% poor low energy response 
Document No. ORAUT-TKBS-0029-6 Revision No. 00 PC-1 Effective Date: 08/03/2006 Page 24 of 42

correction factor for TLD (RTG workers only) and a track fading correction for TLDs. The corrections 

are summarized in Table 6-10.

Table 6-10. Selection of neutron radiation energies and associated 

dose fractions for the RTG building 400 Pinellas Plant processes. 

Facility or 

Location

Process type

Neutron 

energy

Default 

Dose 

Fraction

ICRP 60 

correction 

factor

Building 400 RTG assembly and testing 0-2 MeV 0.25 0.48

Building 400 RTG assembly and testing 2-12 MeV 0.75 0.99

6.3.3

Beta doses were monitored but not routinely recorded. The primary source of beta or non-penetrating 

dose was from the use of two RadiFlow leak testers located in area 109 from about 1963 though 

1994. 

Beta Dose Adjustments

Since beta monitoring for 85Kr started prior to the DOELAP standard release, which used a calibration 

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

compensate for the lower energy of 85Kr relative to that of 90Sr/90Y a correction factor may have been 

used for the Pinellas site based upon the more similar 204Tl energy spectrum. To compensate for the 

energy differences a multiplication factor of about 3.5 (GEND 1985b, PDN 240001210) times the 

reported non-penetrating dose should be used for dosimetry prior to about 1986. From 1986 onwards, 

when DOELAP and NAVLAP standards included 204Tl calibration criteria for 85Kr exposures it is not 

necessary to utilize this correction factor. It is not clear from the Pinellas Plant records whether the 

204Tl energy calibration or equivalent was requested by Pinellas of Landauer or other vendors prior to 

1986.

6.4 MISSED DOSE

There is missed dose for Pinellas Plant workers. The following sections discuss photon and neutron 

missed dose. 100 mrem (95th percentile) dose equivalent will be assigned as a constant for 

unmonitored workers. Missed dose will NOT be assigned. Due to the nature of the site and the 

potential for exposure the 100 mrem constant will be claimant favorable.

6.4.1

Missed photon and beta dose to Pinellas Plant workers could occur for the following reasons: (1) 

there is no recorded dose for workers who did not work in certain operations, and (2) a worker dose 

during a monitoring period was recorded as zero because the dosimeter response was less than the 

MDL. Many workers were not monitored for external dose because such doses would not exceed 

10% of the radiation protection guides. This practice resulted in large numbers of workers not being 

monitored for external radiation exposure from 1957 through 1994. Note, the vast majority of these 

workers may not have worked with or near any radioactive materials or areas since most of the 

Pinellas Plant products did not include the use of radioactive materials or radiation generating 

devices.

Photon and Beta Missed Dose

Missed dose is primarily estimated on dosimeter results n (the number of zero or < MDL values) 

multiplied by MDL/2. The MDL is particularly important during the early years of operation, when 

MDLs were probably higher and the dosimeter exchange rate was monthly rather than quarterly. One 
Document No. ORAUT-TKBS-0029-6 Revision No. 00 PC-1 Effective Date: 08/03/2006 Page 25 of 42

option to estimate a claimant-favorable maximum potential dose is to multiply the MDL by the number 

of zero dose results. This will provide an estimate of the maximum missed dose to the worker. The 

following sections consider missed photon dose for dosimeter results less than the MDL according to 

facility or location, dosimeter type, year, and energy range.

The analysis assumed that unmonitored (i.e., nonradiation) workers did not receive a significant dose 

compared to monitored workers; therefore, assigning a photon dose distribution for each year based 

on the dose received by monitored workers would ensure a claimant-favorable estimate of any 

unmonitored worker 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. 

6.4.1.1 Missed Dose by Facility or Location

Table 6-11 lists potential missed dose by facility or location. Records that identified facilities or 

locations in which specific types of dosimetry were used are not available. Rather, the types of 

dosimeters were assigned to facilities and locations based on the operations that were conducted.

Table 6-11. Photon and beta dosimetry missed dose by facility or location. 

Facility or locationa Period of use Dosimeter MDL (rem)b

Exchange 

frequency

Max. annual missed 

dose (rem)c

Building 100, Areas 107, 108, 109, 

132M, 155, 157/158, 182, 191

Buildings 200

1957- June 1974 film – whole 

body

0.04 Monthly (n=12) 0.24 beta -photons

Building 100, Areas 107, 108, 109, 

132M, 155, 157/158, 182, 191

Buildings 200, 400, 1000

July 1974 –

About 1983d

Landauer 

Type K (betagamma)- 

whole body

0.01 photons

0.04 beta

Monthly (n=12) 0.06 photons

0.24 beta

Building 100, Areas 107, 108, 109, 

132M, 155, 157/158, 182, 191

Buildings 200, 400, 1000

July 1974 – 

About 1983d

Landauer 

Type J – 

whole body

0.01 photons

0.04 mrem beta

Monthly (n=12) 0.06 photons

0.24 beta

Building 100, Areas 107, 109, 114, 

131, 132, 139, 148, 161, 175, 176, 

181, 183, 184, 191, 194, 196; 

Buildings 200; 300; 400; 800; Medical

About 1983d – 

April 1990

Landauer G1 0.01 photons

0.04 beta

Monthly until Jan. 

1990 (n=12)

Quarterly afterward 

(n=4)

0.06 photons (monthly)

0.02 photons (quarterly)

0.24 beta (monthly)

0.08 beta (quarterly)

Building 100, Areas 107, 109, 114, 

131, 132, 139, 148, 161, 175, 176, 

181, 183, 184, 191, 194, 196; 

Buildings 200; 300; 400; 800; Medical

April 1990 –

1997

Landauer Z1 

dosimeter

0.01 photons

0.04 beta

Quarterly (n=4) 0.02 photons

0.08 beta

Building 100, Areas 107, 131, 161, 

175, 181

1957-June 1974 Film badge -

wrist

0.04 Monthly (n=12) 0.24 beta -photons

Building 100, Areas 107, 131, 161, 

175, 181; 400

July 1974 –

About 1983d

Landauer 

Type M (wrist)

0.01 photons

0.04 beta

Monthly (n=12) 0.06 photons

0.24 beta

Building 100, Areas 107, 131, 161, 

175, 181; Building 400

About 1983d –

1990

G5 wrist film 

badge

0.01 photons

0.04 beta

Monthly (n=12) 0.06 photons (monthly)

0.02 photons (quarterly)

0.24 beta (monthly)

0.08 beta (quarterly)

Building 100, Areas 107, 131, 161, 

175, 181; Building 400

1991 – 1997 K5 TLD wrist 

badge

0.01 photons

0.04 beta

Quarterly (n=4) 0.02 photons

0.08 beta

Building 100, Areas 131, 139 1957-June 1974 Film badge -

finger

0.04 Monthly (n=12) 0.24 beta -photons

Building 100, Areas 131, 139; 

Building 400

About 1983d –

1997

U3 TLD (LiF) 

finger ring

0.03 photons

0.04 beta

Monthly until Jan. 

1990 (n=12)

Quarterly 

afterwards (n=4)

0.18 photons (monthly)

0.06 photons (quarterly)

0.24 beta (monthly)

0.08 beta (quarterly)

a. As determined from site operational information.

b. Estimated MDLs for commonly used photon dosimetry (ORAU 2004a).

c. Maximum annual missed dose calculated using the MDL divided by 2 (NIOSH 2002).

d. Dates are approximate. No information could be found to determine when the G1 replaced the Type K and J, the G5 replaced the Type 

M, and use of the U3 began.
Document No. ORAUT-TKBS-0029-6 Revision No. 00 Effective Date: 09/15/2005 Page 26 of 42

This analysis assumed that workers wore finger and wrist dosimetry in Building 400 for the production 

of RTGs due to the handling of plutonium (Weaver 1987).

Unmonitored dose should be assigned to account for external dose that may have received that was 

not measured. Summarized dosimetry data available for the period 1983–1993 indicates that the 

highest annual external dose for an individual at the Pinellas Plant was 0.550 rem. Typical annual 

dosimetry results for personnel monitored between 1957 and 1979 indicated few individuals received 

doses greater than 0.500 rem. Even though doses at the Pinellas Plant were expected to be lower 

than the maximum individual dose, the maximum doses may be assigned based on an assumption of 

0.550 rem to each full or partial year. However, typically over 95% of monitored workers received 

annual doses below 0.100 rem. The only exceptions found to this were for the years 1958 and 1960 

when only 80% and 84% of the monitored population were below 0.100 rem. The data indicates that 

an annual dose of 0.100 rem is representative of the upper 95th percentile dose for the Pinellas Plant.

The unmonitored photon doses would be adjusted by DCF and other applicable factors.

6.4.1.2 Missed Dose by Dosimeter Type and Year

Table 6-11 summarizes the missed photon dose by dosimeter type and year. The MDLs for the 

specific beta/photon dosimeters used until the 1970s are not known; however, the dosimeters would 

be comparable to those used at other AEC/DOE sites. A claimant-favorable assumption of 40 mrem is 

appropriate for these dosimeters (NIOSH 2004a). 

6.4.1.3 Missed Dose by Energy Range

An estimate of the missed dose by energy range is possible based on the predominant radiation 

sources or radionuclides used at the various Plant facilities, primarily lower (<100 keV) energy 

photons from plutonium. The recorded dose from the dosimeter response does not typically provide 

information to estimate discrete energy ranges. It is possible to examine the energy response 

characteristics of the respective multi-element dosimeters, but such an analysis does not recognize 

the substantial uncertainties present in the workplace associated with shielding, radiation scattering, 

and mixed radiation fields.

6.4.2

The possibility of missed neutron dose at the Pinellas Plant could have occurred due to the following:

Neutron Missed Dose

• Track fading in NTA film.

• An MDL of 50 mrem for NTA film (ORAU 2003a).

• Under-response of NTA film to lower energy neutrons, particularly those less than 500 keV, 

including albedo neutrons.

• MDLs as noted in Table 6-13 for TLDs.

• Under-response of TLDs to lower energy neutrons (< 2 MeV) as calculated by GEND Health 

Physics for RTG neutron spectrum.

• Signal fading in TLD neutron dosimeters
Document No. ORAUT-TKBS-0029-6 Revision No. 00 PC-1 Effective Date: 08/03/2006 Page 27 of 42

Table 6-12 summarizes the neutron missed dose by facility or location segregated by the mode of the 

missed dose (i.e., track fading, MDL, and poor energy response).

Table 6-12. Neutron dosimetry missed dose by facility or location. 

Facility or locationa

Period 

of use Dosimeter

Missed dose per badge 

exchange (rem) Exchange frequency

Max. annual 

missed dose 

(rem)

Building 100, Areas 128, 

131, 148, 176, 181, 182, 

183, 184, 191, 194, 196;

Buildings 200; 800, 400

1957 – June 

1978b

NTA track film 0.0335 (track fading)c{1957-

1969}

Monthly (n=12) 0.402

0.05 (MDL)d Monthly (n=12) 0.60

0.03 (poor energy 

response)e {For RTG 1975-

1990 only}

Monthly (n=12) 0.36

Building 100, Areas 128, 

131, 148, 176, 181, 183, 

184, 191, 194, 196;

Buildings 200; 800, 400

July 1978 –

About 1983b

Landauer Type K 0.0335 (track fading)c Monthly (n=12) 0.402

0.04 (MDL)f Monthly (n=12) 0.24

0.03 (poor energy 

response)e {For RTG 1975-

1990 only}

Monthly (n=12) 0.36

Building 100, Areas 128, 

131, 148, 176, 181, 183, 

184, 191, 194, 196;

Buildings 200; 800 

About July 

1978b – 1997

Landauer 

Neutrak E1 

dosimeter

0.02 (MDL)f Monthly until Jan. 1990 (n=12)

Quarterly afterwards (n=4)

0.24

0.08

Building 100, Area 182;

Building 400

October 1979 

– October 

1987

Mound albedo 

dosimeter

0.01 (MDL)d Monthly (n=12) 0.12

0.011 (signal fading)h Monthly (n=12) 0.132

0.0033 (poor energy 

response)f{For RTG 1975-

1990 only}

Monthly (n=12) 0.040

Building 100, Area 182;

Building 400

September

1987 – 1990 

Landauer I8 

Neutrak 

Extended Range 

dosimeter

0.02 (MDL)f Monthly until Jan. 1990 (n=12)

Quarterly afterwards (n=4)

0.24

0.08

0.0066 (energy dose 

response)f{For RTG 1975-

1990 only}

Monthly until Jan. 1990 (n=12)

Quarterly afterwards (n=4)

0.079

0.020

a. As determined from site operational information.

b. Dates are approximate. No documentation could be found to verify the dates for when the E1 replaced the Type J and when the Type J 

replaced the NTA track film. Use of area monitoring for personnel monitoring for neutrons may have occurred until May 1960. The 

missed dose calculation would likely be favorable to claimants. (GEND 1960). 

c. 67% track fading per month as determined in a fade study (GEND 1969). This is multiplied by the MDL of 0.05 rem to obtain the 

amount of missed dose due to track fading per badge exchange.

d. (ORAU 2004b, Table 6-17). 

e. Film NTA poor energy response calculated from the Savannah River Site correction factor of 1.14 neutrons in the energy range of 0.1 

to 2 MeV. This correction would be only applied to RTG exposures (ORAU 2003a).

f. 67% dose response of Neutrak ER or I8 dosimeter to RTG neutrons as determined by GEND health physics group. 1.33 is multiplied 

by the MDL to obtain the missed dose due to poor low-energy response leading to a poor dose response. This correction would be only 

applied to RTG exposures (240001308 – 1987). 

g. Landauer dosimetry specifications from Passmore (2004).

h. Mound dosimetry correction for signal fading (ORAU 2004b).

Personnel monitoring for neutrons may have began in 1960. Not all neutron tube workers were 

required to wear badges (Mevers 2005). Note, not all workers designated as tube workers were tube 

testers. For external dose, only tube testers would have been exposed to neutron radiation. Neutron 

dose from 1957-1959 may have been estimated from area monitors. The missed dose calculation for 

1957-1959 as presented in Table 6-12 would be favorable to claimants. 

6.4.2.1 Missed Dose by Dosimeter Type and Year

Table 6-12 summarizes the MDL to be used for each dosimetry technology and the corresponding 

missed dose segregated by the mode of the missed dose (i.e., track fading, MDL, and poor energy 

response). Little information could be found regarding the MDL for each dosimetry technology 

throughout most of the history of the Pinellas Plant. Therefore, an MDL is proposed for cases in which 
Document No. ORAUT-TKBS-0029-6 Revision No. 00 Effective Date: 09/15/2005 Page 28 of 42

a limit is not known so missed dose can be calculated. The purpose of the proposed MDL values is to 

be claimant-favorable.

6.5 UNCERTAINTY

For measuring film badge doses, the MDLs quoted in the literature ranged from about 10 to 40 mrem 

for beta/photon radiation; it is possible to read a photon dose of 100 mrem to within 15 mrem if the 

exposure involved photons with energies between several hundred keV and several MeV (Morgan 

1961). The estimated standard error in recorded film badge doses from photons of any energy is 

±30%. 

For NTA films, the estimated standard error for the assigned photon dose is within 30%. When 

measuring neutrons with NTA films, the estimated standard error for the assigned neutron dose is 

within 50 mrem. The sensitivity of the polycarbonate neutron dosimeter (Type E1) was 30 ± 15 mrem, 

as measured in 1978, with a minimum detection limit of 20 mrem (Holliday 1978).
Document No. ORAUT-TKBS-0029-6 Revision No. 00 PC-1 Effective Date: 08/03/2006 Page 29 of 42

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Pinellas Area Office, St. Petersburg, Florida, March 28,.

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fax date October 27.

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Topics in Radiation Dosimetry, International Atomic Energy Agency, Vienna, Austria, pp. 3-23.

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Neutron Radiation, Report 38, Bethesda, Maryland.

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2004.

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GLOSSARY

absorbed dose, D

Amount of energy imparted by radiation to unit mass of absorbing material (100 ergs per 

gram), including tissue. The unit used prior to the use of the International System of metric 

units (SI) is the rad; the SI unit is the gray.

accreditation

In relation to this document, recognition that a dosimeter system has passed the performance 

criteria of the DOE Laboratory Accreditation Program (DOELAP) standard (DOE 1986) in 

specified irradiation categories.

accuracy

If a series of measurements has small systematic errors, they are said to have high accuracy. 

The accuracy is represented by the bias.

albedo dosimeter

A TLD device that measures the thermal, intermediate, and fast neutrons that are scattered 

and moderated by the body from an incident fast neutron flux.

Atomic Energy Commission

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

Manhattan Engineering District (MED) and a predecessor to DOE.

backscatter

Deflection of radiation by scattering processes through angles greater than 90 degrees, with 

respect to the original direction of motion.

beta particle

A charged particle of very small mass emitted spontaneously from the nuclei of certain 

radioactive elements. Most (if not all) of the direct fission products emit (negative) beta 

particles. Physically, the beta particle is identical with an electron moving at high velocity.

claimant-favorable

Refers to the process of estimation based on technical considerations of the parameters 

significant to dose such that there is no underestimation of the estimated dose.

curie

A special unit of activity. One curie exactly equals 3.7 × 1010 nuclear transitions per second.

densitometer

Instrument that has a photocell to determine the degree of darkening of developed 

photographic film.

dose equivalent (H)

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

special unit is the rem. When D is expressed in Gy, H is in Sieverts (Sv). (1 Sv = 100 rem.)

DOELAP

The DOE Laboratory Accreditation Program accredits DOE site dosimetry programs based on 

performance testing and onsite reviews performed on a 2-year cycle.
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dose equivalent index

For many years the dose equivalent used to calibrate neutron sources that were used in turn 

to calibrate neutron dosimeters; a concept of summing the maximum dose equivalent 

delivered in the International Commission on Radiation Units and Measurements sphere at 

any depth for the respective neutron energies even though the maximum dose occurred at 

different depths.

dosimeter

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

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

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

dosimeter, neutron film dosimeter, thermoluminescent dosimeter.)

dosimetry system

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

and/or extremities. This includes the fabrication, assignment, and processing of the dosimeters 

as well as interpretation and documentation of the results.

error

Term used to express the difference between the estimated and "true" value. Error can also be 

used to refer to the estimated uncertainty.

exchange period (frequency)

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

exposure

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

ionization produced by gamma (or X-) rays in air. 

extremity

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

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

toes.

fast neutron

Neutron of energy between 10 keV and 10 MeV.

film

Generally means a "film packet" that contains one or more pieces of film in a light-tight 

wrapping. The film when developed has an image caused by radiation that can be measured 

using an optical densitometer. 

film dosimeter

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

filter

Material used to adjust radiation response of a dosimeter to provide an improved tissue 

equivalent or dose response.
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gamma rays

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

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

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

not originate in the nucleus. 

gray (Gy)

The SI unit of absorbed dose (1 Gy = 100 rad).

intermediate energy neutron

Neutron of energy between 0.5 ev (assumed to be 0.4 ev because of cadmium cutoff in 

neutron response) and 10 keV.

ionizing radiation

Electromagnetic radiation (consisting of photons) or particulate radiation (consisting of 

electrons, neutrons, protons, etc.) capable of producing charged particles through interactions 

with matter.

isotopes

Forms of the same element having identical chemical properties but differing in their atomic 

masses. Isotopes of a given element all have the same number or protons in the nucleus but 

different numbers of neutrons. Some isotopes of an element can be radioactive.

kilo-electron volt (keV)

An amount of energy equal to 1,000 electron-volts.

luminescence

The emission of light from a material as a result of some excitation.

Minimum Detection Level (MDL)

Often confused because the statistical parameters necessary to its calculation are not 

explicitly defined. Nonetheless, often assumed to be the level at which a dose is detected at 

the two-sigma level (i.e., 95% of the time). The MDL should not be confused with the minimum 

recorded dose. 

minimum recorded dose

Based on a policy decision, the minimum dose level that is routinely recorded. A closely 

related concept is the dose recording interval. 

million-electron volt (MeV)

An amount of energy equal to 1,000,000 electron-volts.

neutron

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

neutron, fast

Neutrons with energy equal or greater than 10 keV.

neutron, intermediate

Neutrons with energy between 0.4 eV and 10 keV.
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neutron, thermal

Strictly, neutrons in thermal equilibrium with surroundings. In general, neutrons with energy 

less than the cadmium cutoff at about 0.4 eV.

Personal Dose Equivalent, Hp(d)

Radiation quantity recommended for use as the operational quantity to be recorded for 

radiological protection purposes by the International Commission on Radiological Units and 

Measurements. Represented by Hp(d), where d identifies the depth (in millimeters) and 

represents the point of reference for dose in tissue. For weakly penetrating radiation of 

significance to skin dose, d = 0.07 mm and is noted as Hp(0.07). For penetrating radiation of 

significance to whole-body dose, d = 10 mm and is noted as Hp(10). 

photon

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

quality factor, Q

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

rad

A unit of absorbed dose equal to the absorption of 100 ergs per gram of absorbing material, 

such as body tissue.

radiation

One or more of beta, neutron, and photon radiation. 

radiation monitoring

Routine measurements and the estimation of the dose equivalent for determining and 

controlling the dose received by workers.

radioactivity

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

neutrons from unstable nuclei

radioisotopic thermoelectric generator

A radioisotope thermoelectric generator (RTG) is a very simple electrical generator that 

obtains its power from passive radioactive decay. Such a generator uses the fact that 

radioactive materials (such as 238Pu/239Pu) generate heat as they decay. The heat used is 

converted into electricity by an array of thermocouples. 

random errors

When a given measurement is repeated, the resulting values, in general, do not agree exactly. 

The causes of the disagreement between individual values must also be causes of their 

differing from the "true" value. Errors resulting from these causes are random errors.

rem

A unit of dose equivalent, which is equal to the product of the number of rads absorbed and 

the quality factor.
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roentgen

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

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

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

scattering

The diversion of radiation from its original path as a result of interactions with atoms between 

the source of the radiation and a point at some distance away. Scattered radiations are 

typically changed in direction and of lower energy than the original radiation.

shielding

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

personnel or materials from radiation.

Sievert (Sv)

The SI unit for dose equivalent. (1 Sv = 100 rem.)

skin dose

Absorbed dose at a tissue depth of 7 mg/cm2

. 

systematic errors

When a given measurement is repeated and the resulting values all differ from the "true" value 

by the same amount, the errors are systematic.

thermal neutron

Strictly, neutrons in thermal equilibrium with surroundings. In general, neutrons of energy less 

than the cadmium cutoff of about 0.4 ev.

tissue equivalent

Used to imply that the radiation response characteristics of the material being irradiated are 

equivalent to tissue. Achieving a tissue-equivalent response is typically an important 

consideration in the design and fabrication of radiation measuring instruments and dosimeters.

thermoluminescent

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

thermoluminescent dosimeter (TLD)

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

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

chips are sometimes called crystals.

whole-body dose

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

); also 

used to refer to the dose recorded.

X-ray

Ionizing electromagnetic radiation of extranuclear origin.
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ATTACHMENT A 

EXAMPLES OF PINELLAS PLANT EXPOSURE RECORDS
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