ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 39

DOE Review Release 07/27/2011

Document Title:

Pinellas Plant – Occupational Internal Dose 

Document Number: ORAUT-TKBS-0029-5 

Revision: 01

Effective Date: 04/01/2011

Type of Document: TBD

Supersedes: Revision 00

Subject Expert(s): Brian P. Gleckler, and Mutty M. Sharfi

Approval: Signature on File Approval Date: 03/29/2011

Brian P. Gleckler, Document Owner

Concurrence: Signature on File Concurrence Date: 03/29/2011

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 03/29/2011

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 03/29/2011

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 04/01/2011

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-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 2 of 39

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

10/07/2005 00 New technical basis document for the Pinellas Plant – Occupational 

Internal Dose. First approved issue. Training required: As 

determined by the Task Manager. Initiated by Mark D. Notich.

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

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

0015. The changes that potentially affect the assessed doses for 

workers include: 1) the addition of an approach for assessing 

potential exposures to insoluble forms of tritium and 2) the 

modification of some of the default tritium urinalysis MDCs and/or

reporting levels. In addition, a number of editorial changes were also 

made to this document. Incorporates formal internal and NIOSH 

review comments. Constitutes a total rewrite of the document. 

Training required: As determined by the Objective Manager. 

Initiated by Brian P. Gleckler.
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 3 of 39

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations ...............................................................................................................5

5.0 Occupational Internal Dose......................................................................................................7

5.1 Introduction ..............................................................................................................................7

5.1.1 Overview ......................................................................................................................8

5.1.2 Purpose........................................................................................................................8

5.1.3 Scope...........................................................................................................................8

5.2 Radioactive Source Term and Lung Absorption Types............................................................. 9

5.2.1 Tritium ..........................................................................................................................9

5.2.2 Plutonium ...................................................................................................................10

5.2.3 Uranium......................................................................................................................11

5.2.4 Nickel-63 ....................................................................................................................12

5.2.5 Carbon-14 ..................................................................................................................12

5.2.6 Krypton-85..................................................................................................................13

5.2.7 Miscellaneous Radionuclides......................................................................................13

5.3 Historical Monitoring Practices ............................................................................................... 13

5.3.1 Contamination Monitoring Practices ...........................................................................13

5.3.1.1 Tritium Contamination Monitoring................................................................. 13

5.3.1.2 Plutonium Contamination Monitoring............................................................ 14

5.3.2 Air Sampling Practices................................................................................................14

5.3.2.1 Tritium Air Sampling..................................................................................... 14

5.3.2.2 Plutonium Air Sampling................................................................................ 15

5.3.3 Bioassay Monitoring Practices....................................................................................16

5.3.3.1 Records Interpretation.................................................................................. 16

5.3.3.2 Tritium Bioassay Monitoring ......................................................................... 16

5.3.3.3 Plutonium Bioassay Monitoring .................................................................... 18

5.4 In Vitro Bioassay Detection Levels and Analysis Methods...................................................... 19

5.4.1 Urinalysis Detection Levels.........................................................................................19

5.4.1.1 Tritium Urinalysis Detection Levels .............................................................. 19

5.4.1.2 Plutonium Urinalysis Detection Levels.......................................................... 20

5.4.2 Analysis Methods for In Vitro Urine Samples ..............................................................20

5.4.2.1 Analysis Methods for Tritium ........................................................................ 20

5.4.2.2 Analysis Methods for Plutonium ................................................................... 21

5.5 In Vivo Minimum Detectable Activities and Analysis Methods ................................................ 22

5.6 Interferences and Uncertainties.............................................................................................. 22

5.6.1 Tritium Bioassay.........................................................................................................22

5.6.2 Plutonium Bioassay ....................................................................................................23

5.7 Assessment of Intakes And Doses......................................................................................... 23

5.7.1 Workers Monitored for Soluble Tritium Exposure........................................................23

5.7.1.1 Assessment of Soluble Tritium Exposures ................................................... 23
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 4 of 39

5.7.1.2 Assessment of Potential Unmonitored Exposures to Insoluble 

Tritium.......................................................................................................... 23

5.7.2 Workers Not Monitored for Soluble Tritium Exposure .................................................24

5.7.3 Workers Monitored for Plutonium Exposure................................................................24

5.7.4 Workers Not Monitored for Plutonium Exposure .........................................................25

5.8 Incident History ......................................................................................................................25

5.9 Attributions and Annotations .................................................................................................. 28

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

Glossary ............................................................................................................................................35

LIST OF TABLES

TABLE TITLE

5-1 Potential tritium radioactive material use areas ......................................................................

PAGE

10

5-2 Plutonium radioactive material management areas ................................................................ 10

5-3 Default tritium urinalysis MDCs and reporting levels............................................................... 19

5-4 Insoluble tritium intake rates................................................................................................... 24

5-5 Chronology of unusual events and significant activities in relation to internal 

dosimetry ...............................................................................................................................26
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 5 of 39

ACRONYMS AND ABBREVIATIONS

AMAD activity median aerodynamic diameter

AEC U.S. Atomic Energy Commission

CF calibration factor

CFR Code of Federal Regulations

cm centimeter

cpm counts per minute

DIL derived investigation level

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

DU depleted uranium

EDE effective dose equivalent

EEOICPA Energy Employees Occupational Illness Compensation Program Act

g gram

GE General Electric Company

GEND GE Neutron Devices 

GENDD GE Neutron Devices Department 

GEPP GE Pinellas Plant

GEXF GE X-ray Division in Florida

GEXM GE X-ray Division in Milwaukee, Wisconsin

HEPA high-efficiency particulate air

hr hour

ICRP International Commission on Radiological Protection

keV kiloelectron-volt, 1,000 electron-volts

HT tritium gas

HTO tritiated oxide

L liter

LANL Los Alamos National Laboratory

LOD limit of detection

LSC liquid scintillation counting

MDA minimum detectable activity

MDC minimum detectable concentration

MDL minimum detection limit

MeV megaelectron-volt, 1 million electron-volts

min minute

mL milliliter

mR milliroentgen

MT metal tritide

mrem millirem

NIOSH National Institute for Occupational Safety and Health

NIST National Institute of Standards and Technology
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 6 of 39

OBT organically bound tritium

POC probability of causation

POPOP 1,4-bis[5-phenyloxazol-2-yl]benzene (a secondary scintillator)

PPO 2,5-diphenyloxazole (a primary scintillator)

RBE relative biological effectiveness

ROI region of interest

RTG radioisotopically-powered thermoelectric generator

SRDB Ref ID Site Research Database Reference Identification (number)

T tritium

T2 tritium gas

T2O tritium oxide

TBD technical basis document

U.S.C. United States Code

μCi microcurie

μm micrometer

§ section or sections
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 7 of 39

5.0 OCCUPATIONAL INTERNAL DOSE

5.1 INTRODUCTION

 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

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

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

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

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

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

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

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

determined in accordance with the POC [probability of causation1

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

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

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

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

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

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

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

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

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

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

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

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

occupationally derived (NIOSH 2010):

] guidelines established under 

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

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

DOE employees with a covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2010).

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

conventional structures

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

reasons

 

1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 8 of 39

5.1.1

This TBD is Part 5 of the Pinellas Plant’s Site Profile. A site profile provides a summary of information 

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

Overview

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

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

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

Plant. 

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

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

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

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

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

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

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

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

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

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

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 ultraclean, high-vacuum 

technology; hermetic seals between glass, ceramic, glass-ceramic, and metal materials; and highvoltage generation and measurement technology. The Plant manufactured only neutron generators 

for its first 10 years of operation. It later manufactured other products including neutron detectors, 

radioisotopic thermoelectric generators (RTGs), high-vacuum switch tubes, specialty capacitors, and 

specialty batteries (Weaver 1990). As part of its program to promote commercial uses of the site, 

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

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

5.1.2

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

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

internal occupational radiation dose for EEOICPA claims. 

Purpose

5.1.3

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

in accordance with the guidelines described in Internal Dose Reconstruction Implementation 

Guideline (NIOSH 2002). NIOSH considers the available data and methods for performing internal 

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

Pinellas Plant throughout its entire history. 

Scope
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 9 of 39

5.2 RADIOACTIVE SOURCE TERM AND LUNG ABSORPTION TYPES

5.2.1

Tritium (also denoted as T, H-3, and 3

H) is a hydrogen atom with two neutrons. It is the heaviest of 

the three isotopes of hydrogen (protium [1

H], deuterium [2H], and tritium [3H]) and is the only 

radioactive hydrogen isotope. Tritium is a low-energy, beta-emitting radionuclide with a half-life of 

12.28 years (Kocher 1981). The average and maximum beta particle energies are 5.7 keV and 18.6 

keV, respectively (Kocher 1981). Between 1957 and 1993, annual tritium inventories at the Pinellas 

Plant ranged from 5.44 g (5.24 × 104 Ci) to 53.27 g (5.14 × 105 Ci) (Biedermann 1994). Tritium is not 

considered to be an external radiation hazard because the beta particles being emitted have too low 

of an energy to penetrate human skin. However, inside the body its radiation can cause damage to 

tissues and organs. 

Tritium

Four primary forms of tritium were present at the Pinellas Plant: tritiated water (HTO), tritium gas (HT 

or T2), organically bound tritium (OBT), and metal tritides (MTs). One of the most common forms of 

organically bound tritium at the Pinellas Plant was contaminated pump oils. The MTs (primarily 

scandium tritide, erbium tritide, and titanium tritide) formed during production processes could have 

been released in the work environment as particulate aerosols. The gas was allowed to react with 

metal surfaces, thin metal coatings, and metal powders for various purposes. Powders were normally 

contained with vacuum systems, and metal systems normally remained intact (Burkhart 1995a, p. 2). 

Based on the available information about the Plant’s metal tritide uses, only a small portion of the 

worker population had the potential to be exposed to a dispersible form of a metal tritide, and those 

scenarios were typically limited to accidents involving metal tritides. Titanium tritide was also used in 

the original tritium storage beds at the Pinellas Plant (Burkhart 1990a). In the original tritium storage 

beds, the titanium tritide was sealed inside a glass cylinder (Burkhart 1990a). Because the glass 

cylinders for the tritium storage beds broke on occasion, resulting in high levels of contamination, the 

original glass tritium storage beds were replaced with stainless-steel tritium storage beds in 1968 

(Phillips 1975). The stainless-steel tritium storage beds utilized a different MT, DU tritide (Burkhart 

1990a; Eichman1979; Phillips 1975). Because DU tritides were sealed within a stainless-steel 

construction, and because there are no known incidents of DU tritide contamination at the Pinellas 

Plant, exposures to DU and DU tritide were unlikely. Table 5-5, which is provided later in this 

document, includes several incidents involving the various tritium compounds that were present at the 

Pinellas Plant. 

The potential tritium use areas at the Pinellas Plant are summarized in Table 5-1. There have been a 

number of tritium contamination incidents, as discussed in Section 5.8. 

Both soluble and insoluble tritium compounds were present at the Pinellas Plant. Tritium compounds 

with the lung absorption properties of type M or S are considered to be insoluble tritium compounds. 

All other forms of tritium are considered to be soluble tritium compounds. 
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 10 of 39

Table 5-1. Potential tritium radioactive material use areas. 

Location Name Activity

107 Tube Assembly Vacuum tube manufacturing and coating

108 Tube Exhaust and Test Vacuum tube evacuation and testing

109 Product Analysis Magnetic and radioactive gas leak checking

128 Tube Test

131 Final Tube Test

132M Fan Room Stack effluent control and tritium recovery

157/158 Gas Analysis Laboratory Hydrogen isotope analysis

182-C Tube Assembly Vacuum tube development and testing

182-G Tube Exhaust Vacuum tube development and testing

183 General Development

191 CPE Hood Room Ceramic product testing facility

Radioanalytical Laboratory Radiological Laboratory, tritium recovery

194 Engineering Environmental Testing

200 Test areas

800 Accelerator and Calibration

1000 Waste Storage

Source: DOE 1995, p. 5-3; Weaver 1993. 

5.2.2

In terms of radioactive decay, 238Pu and 239Pu are alpha and X-ray emitting radionuclides with halflives of 87.75 years and 24,131 years, respectively (Kocher 1981). The alpha particle emissions from 

these two radionuclides are the primary concerns in regards to internal dose. The photon (gamma ray 

and X-ray) and neutron radiation emitted from the sources containing plutonium are the primary 

concerns in regards to external dose. Gamma rays and neutrons are predominately emitted from the 

spontaneous fissions of plutonium, alpha-neutron reactions, and photon-neutron reactions.

Plutonium

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

in January 1957 (Author unknown undated a). The source was used for calibrating health physics 

monitoring equipment (Author unknown c. 1993a). Based on the available information regarding this 

source and its use, it was most likely an encapsulated plutonium-beryllium (Pu-Be) neutron source. 

The triply encapsulated plutonium oxide (

238PuO2) heat sources that were used for the radioisotopic 

thermoelectric generators (RTGs) did not start arriving at the Pinellas Plant until November 1975 

(Author unknown undated a). There were two different types of 238PuO2 heat sources, 8.75 g sources 

and 10 g sources (GE 1982a). With the exception of 1975, no information was found regarding the 

annual inventories of 238PuO2 heat sources. In November 1975, the site received seven 238PuO2 heat 

sources (Author unknown undated a). By February 1991, all plutonium sources, with the exception of 

calorimeter sources and small instrument calibration check sources, were removed from the Pinellas 

Plant (MMSC 1992). 

The potential plutonium use areas at the Pinellas Plant are summarized in Table 5-2. 

Table 5-2. Plutonium radioactive material 

management areas. 

Location Name

400 RTG area

200 Test areas

800 Accelerator

1000 Waste storage
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 11 of 39

Based on the 1982 version of the Safety Analysis Report (SAR) for the RTG Facility, shipments of the 

238PuO2 heat sources were delivered directly to Building 400, where the RTG Facility is located (GE 

1982a). When the shipping packages were to be opened, they were moved from the vault room to 

the source inspection hood where they were opened and surveyed for contamination (GE 1982a). 

Sources confirmed to be free of contamination were placed in a source storage container and 

returned to the vault room. If the unpacking survey showed contamination levels greater than 200 

dpm, the source was to be immediately repackaged and returned to the supplier (GE 1982a; Huffman 

1979). If the survey showed detectable contamination that was below 200 dpm, an effort would be 

made to decontaminate the source in accordance with procedures.

Plutonium sources were considered to be free of significant contamination if the removable 

contamination on a swipe of the entire source surface area was less than twice the statistical counting 

error associated with a 5-minute count and a 95% confidence level (GE 1982a). These criteria 

resulted in a control level of ≤ 6 dpm (GE 1982a). A letter dated January 1979 indicated that the 

occurrence of 238PuO2 heat sources with contamination levels exceeding the limit of detection (LOD) 

was only about one in every 70 sources received or less than 6 per year (Huffman 1979). As of 1982, 

the SAR indicates that the circumstance of a contaminated 238PuO2 heat sources needing to be 

returned to the supplier had not occurred. The SAR also states that the probability of a 238PuO2 heat 

source leaking, although possible, is so small that it can be assumed that it will not occur. In another 

section of the SAR, it is stated that the “Gross failure of heat source encapsulation is not considered 

to be a credible event.” The 1982 version of the SAR also states that “there is not, nor has there ever 

been, any plutonium contamination inside the facility nor released to the environment”, which was 

reiterated in a 1989 memorandum (GE 1982a, Weaver 1989a). 

Based on the information provided in the RTG Facility’s SAR, plutonium intakes were extremely 

unlikely at the Pinellas Plant and the only probable plutonium intake scenario is a receipt inspection 

scenario involving a contaminated 238PuO2 heat source. However, any potential intakes attributable to 

such a scenario would have been limited, since no contamination levels exceeding the 200 dpm limit 

appear to have ever been found. 

5.2.3

Depleted and natural uranium, which consist of 234U, 235U, 238U and some of the radioactive progeny 

for these radionuclides, were present at the Pinellas Plant. In terms of radioactive decay, the uranium 

isotopes emit alpha particles and X-rays. However, some of the radioactive progeny emit beta 

particles and gamma rays.

Uranium

The major use of DU was for the tritium storage beds that were first used in 1968 (Phillips 1975). Fifty 

grams of DU metal was used for the particulate uranium metal tritide in each of the tritium storage 

beds (Ward 1973, p. 29). Because the uranium in the tritium storage beds was sealed in stainless 

steel canisters, the uranium was considered to be a containerized source and would have posed little 

to no internal dose hazard. There was no indication that the uranium ever leaked from the storage 

beds at the Pinellas Plant. Given that particulate uranium metal is pyrophoric, any uranium metal 

leaking from the tritium storage beds would have ignited and resulted in a uranium fire incident at the 

site. Of the reported incidents for the Pinellas Plant, none were uranium release or uranium fire 

incidents. The depleted uranium (mainly 238U) inside the tritium storage beds presents no significant 

external radiation hazard, due to the low specific activity and the nonpenetrating radiation emitted. 

The major use of natural uranium was the use of borosilicate glass that was doped with natural 

uranium (1.5% by weight) in the form of U3O8 (Weaver 1992). Because the uranium would have been 

encapsulated in the glass prior to its arrival at the Pinellas Plant, the glass was considered to be a 

sealed source and would have posed little to no internal dose hazard. 
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5.2.4

Nickel-63 is a low-energy, beta-emitting radionuclide with a half-life of 100.1 years. The average and

maximum beta particle energies are 17.13 keV and 65.87 keV, respectively (Kocher 1981).

Nickel-63

The information regarding the Pinellas Plant’s use of 63Ni, includes documents from the GE X-ray 

Division (a.k.a. GEXM) site. Because many of the same nuclear weapons-related activities were 

performed at the Pinellas Plant and GEXM sites, and because many items were obtained from the 

same vendors, use of 63Ni and the form that it was obtained in were likely the same for both sites.

Nickel-63 was electroplated onto a nickel mesh inside a sealed glass tube (a krytron) by U.S. Radium, 

and averaged 0.3 μCi per spark gap (Weaver ca. 1995; Jech 1963). Krytrons are cold-cathode, gasfilled tubes intended for use as very high-speed switches, which have been used for igniting the 

exploding-bridge wire detonators and slapper detonators in nuclear weapons. The 63Ni is used in 

conjunction with the keep-alive electrode, where the beta particles being emitted by the 63Ni make the 

ionization inside the krytron easier. The available information indicates that the Pinellas Plant was not 

involved with the process of electroplating the 63Ni to the keep-alive electrodes for the spark gap 

bodies, and only received the spark gap bodies containing the electroplated 63Ni from U.S. Radium. It

is not known if the electrodes plated with 63Ni were already sealed in the glass tubes before they 

arrived at the Pinellas Plant or if the Pinellas Plant was creating and sealing the electrodes in the 

glass tubes. Given that one of the Pinellas Plant’s areas of expertise was glass formulation, the 

Pinellas Plant was likely sealing the electroplated 63Ni electrodes into the glass tubes to create the 

krytron.

No internal dose monitoring was conducted for 63Ni based on the operations with devices and survey 

data of work areas and parts (Weaver ca. 1995). Pinellas Plant information indicates that hypothetical 

worst-case doses were calculated and placed in an “Internal dosimetry technical notes file” (Weaver 

ca. 1995). However, this file has not been found. Given that each spark gap only contained 

approximately 0.3 μCi (11,100 Bq) of 63Ni and given that the worst-case organ dose coefficient for 63Ni 

is 5.6 × 10-9 Sv/Bq (2.07 × 104 rem/Ci) (ORDOSE 2003), inhaling the total radioactivity in a single 

spark gap body would only result in a maximum committed (50-year) organ dose of approximately 6.2

mrem. Therefore, it is unlikely that any workers at the Pinellas Plant received a significant internal 

dose from 63Ni, and potential 63Ni exposures do not need to be assessed for Pinellas Plant workers. 

As a result, 63Ni is not discussed any further in this TBD. 

5.2.5

Carbon-14 is a low-energy, beta-emitting radionuclide with a half-life of 5,730 years. The average and 

maximum beta particle energies are 49.47 keV and 156.48 keV, respectively (Kocher 1981).

Carbon-14

The use of 14C at the Pinellas Plant is only indicated in the gaseous effluent release reports (GE 1980, 

1981, 1982b, 1983, 1984a) and in an environmental assessment (DOE 1983). The gaseous effluent 

release reports indicate that 14C was used between 1979 and 1983 (GE 1980, 1981, 1982b, 1983, 

1984a). Based on the reported gaseous effluent releases for those years, 14C was used in much 

smaller quantities than tritium. A comparison of the annual quantities of gaseous effluents released 

indicates that the curies of tritium being processed were over 100,000 times greater than the curies of 

14C being processed. A 1983 environmental assessment indicated that small quantities of 14C 

labeled-solvents were used in a laboratory testing operation (DOE 1983). No other documentation 

was found to indicate whether or not there were any other uses of 14C. No documentation was found 

that indicates what chemical forms of 14C were used. Given that 14C use was much less than tritium 

use at the Pinellas Plant, and given that the worst-case organ dose coefficients for 3

H and 14C in the 

Radiological Toolbox computer program (ORDOSE 2003) are within an order of magnitude of each 

other, it is unlikely that 14C was a significant internal dose concern at the Pinellas Plant. Therefore, 
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 13 of 39

internal doses due to 14C exposures do not need to be assessed for Pinellas Plant workers unless 14C 

exposure information is provided in the worker’s dosimetry records. As a result, 14C is not discussed 

any further in this TBD. 

5.2.6

Because 85Kr is a noble gas, it is not a significant internal dose concern. Therefore, internal doses 

due to 85Kr exposures do not need to be assessed, and 85Kr is not discussed any further in this TBD. 

Krypton-85

5.2.7

A wide variety of other radionuclides were used at the Pinellas Plant; however, the uses of these 

radionuclides were mostly limited to sealed and plated check sources, static meter sources, explosive 

meter sources, heat sources, calibration sources, thickness gauges, gas chromatograph sources, dew 

point measurement sources, and static eliminator sources (Author unknown undated b). Even though 

some of these sources contained significant quantities of radioactivity (Author unknown undated b), 

they were not considered to be potential sources for radionuclide intakes, unless a specific worker 

was involved in an incident where an intake pathway was created for one of these sources. 

Therefore, intakes and internal doses for other miscellaneous radionuclides do not normally need to 

be evaluated for Pinellas Plant workers, and these radionuclides are not discussed further in this TBD. 

Any potential intakes of radioactivity and subsequent doses due to an incident involving one of these 

radioactive sources will need to be evaluated on a case-by-case basis. 

Miscellaneous Radionuclides

5.3 HISTORICAL MONITORING PRACTICES

The Pinellas Plant internal dosimetry program started with site operations in 1957. Contamination 

monitoring, air sampling, and bioassay monitoring were the three primary types of monitoring used at 

the Pinellas Plant to detect potential intakes of radioactive materials. 

5.3.1

Tritium and plutonium were the only radioactive materials that contamination monitoring was routinely 

performed for in the production areas. Outside of the production areas at the Pinellas Plant, 

contamination monitoring might have been performed for other radioactive materials; however, no 

documentation regarding other potential types of contamination monitoring has been found. 

Contamination Monitoring Practices

5.3.1.1 Tritium Contamination Monitoring

Contamination monitoring for tritium was performed on a routine basis from the beginning of 

operations at the Pinellas Plant (GE 1957–1973). Work areas and personnel were checked for 

contamination on a routine basis. Any significant personnel contamination that could have gone 

undetected from contamination surveys would most likely have been identified through the tritium 

bioassay program.

The 1957 through 1973 monthly health physics reports (GE 1957–1973) indicate that as early as 

1959, areas greater than 2 × 10-5 μCi/in2 (688 dpm/100 cm2

) are recommended for decontamination. 

In a 1969 health physics report, the control limit was reported as 440 dpm/100 cm2 (GE 1957–1973). 

A circa 1966 smear survey procedure indicated that the uncontrolled area limit for tritium was 220 

dpm/100 cm2 and the controlled area limit was 440 dpm/100 cm2 (GE ca. 1966). The monthly health 

physics reports also provide information on the maximum surface contamination levels. Between 

1957 and 1973, the highest surface contamination level reported was in 1970, 4.4 × 106 dpm/100 cm2

(10,000 times the control limit) (GE 1957–1973). The next highest value was reported in 1959, 

1.4 × 106 dpm/100 cm2 (3.3 × 10-2 μCi/in2

) (GE 1957–1973). The majority of the rest of the annual 
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 14 of 39

maximum surface contamination levels reported are at least an order of magnitude lower than these 

two. 

It was known early that wet swipes for contamination were more efficient than dry swipes. The water 

geometry for counting was used with the liquid scintillation counter for contamination monitoring. This 

same water geometry was used for HTO until about 1971 for the urine bioassays as well (GE 1967–

1982). The procedure for collecting the smears involved pipetting 5 mL of distilled water containing 

the cotton ball swipe, taking the wet cotton ball to the area, and swiping the area of concern (usually 

of 100 to 1,000 cm2

). Upon return to the counting lab, 10 mL of water was added to the swipe 

container, and the sample was filtered through Whatman #1 filter paper. Two-tenths of a milliliter of 

sample was added to 16.5 mL of cocktail solution, and the sample was counted for 5 minutes. The 

minimum detectable concentration (MDC) was 0.002 μCi/area swiped (GE 1967–1982). 

By 1989, contamination monitoring occurred daily in what were considered contamination areas. This 

included Laboratories 158B, 182, and 108. Weekly surveys included Laboratories 158B, 157, 182, 

182G, 108, 132B, 109, 128, and 131. Monthly surveys included the “Pure Zone”; Laboratories 107, 

114 (X-ray), 138, 161, 158A, and 191-N (CPE laboratory); hallways; Buildings 200, 400, 800, and 

1000; and the eating areas. Other areas were surveyed less frequently (GE 1989a; LMSC 1995).

5.3.1.2 Plutonium Contamination Monitoring

In the RTG areas, the Pinellas Plant routinely monitored the work areas for possible plutonium 

contamination from the beginning of plutonium source use for the RTGs in 1975. This occurred 

despite the fact that the sources were triply encapsulated and were held to more stringent standards 

than other Federal regulatory agencies required for sealed sources of radiation. 

Areas were checked for contamination with a portable alpha meter, a PAC 4S proportional detector. 

Because this detector was also sensitive to gamma radiation, users were instructed to place a sheet 

of paper over the detector to absorb the alphas to positively indicate the presence of alpha radiation. 

As indicated in Section 5.2.2, no areas were ever found to have plutonium contamination at the

Pinellas Plant.

5.3.2

Tritium and plutonium were the only radioactive materials that air sampling was routinely performed 

for in work areas. Nonroutine air sampling might have been performed for other radioactive materials; 

however, no documentation regarding other potential types of air sampling has been found. 

Air Sampling Practices

5.3.2.1 Tritium Air Sampling

Monitoring for airborne radioactivity was performed on a routine basis from the beginning of 

operations at the Pinellas Plant (GE 1957–1973). Because tritium was the only known source of 

contamination at the plant during 1957 and because this monitoring appears to have been performed 

using a Kanne ion chamber with a glass wool filter on the intake (GE 1957–1973, page 2), the early 

airborne radioactivity monitoring is assumed to have been limited to gaseous forms of tritium. In 

1957, the maximum permissible concentration for airborne radioactivity was 7 × 10-5 μCi/mL (GE 

1957–1973, p. 5).

Fixed-room monitors were located in all areas where there was a potential for release of tritium. The 

monitoring systems consisted of a 22 L Kanne ionization chamber that was connected to a 

picoammeter and an alarm panel. In 1973, there were 40 sampling ports and 20 monitors. The air 

blowers had backups and an alarm system if they were not working properly. Some Kanne chamber 
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 15 of 39

systems had multiple or sequential sampling systems when more than two ports were attached. This 

allowed sequential sampling for 20-second intervals for each port. A Kanne ionization chamber 

monitor was capable of detecting tritium below the U.S. Atomic Energy Commission (AEC) 40-hr level 

of 2 × 10-5 μCi/mL, which is the same as the current derived air concentration agreed to by DOE and 

the U.S. Nuclear Regulatory Commission (Author unknown 1987; Ward 1973). The MDC has been 

calculated to be as low as 1.3 × 10-5 μCi/mL (Ward 1971). The alarm was set at about four times the 

AEC/DOE/NRC limit of 2 × 10-5 μCi/mL for tritium oxide (i.e., at about 80 μCi/m3 or 8 × 10-5 μCi/mL). 

The operation of each fixed monitoring system was response checked monthly using a small gamma 

source and hand-held smoke generator. Areas monitored included Laboratories 108, 109, 157, 158, 

182, and 132. Pinellas Plant health physicists decided whether chart recorders were to be used, such 

as in cases where suspected higher-than-normal HTO concentrations were anticipated. 

Portable tritium gas monitors were used as temporary monitors in areas where fixed -room monitor 

probes were not located (Author unknown 1987; Ward 1973). Portable samplers using silica gel 

collection media, or silica gel stations were also set up in some areas. 

5.3.2.2 Plutonium Air Sampling

Real-time air monitoring for 238Pu and 239Pu employed Eberline models alpha-2 or alpha-3 room 

monitors in Building 400 (eight locations) using Whatman microfibre #934-AH glass or # 41 ashless 

5.5-cm-diameter filters. The following areas were monitored (Burkhart and Richardson 1986):

• The receiving room where newly arrived heat sources were unpacked; 

• The gloveboxes where the heat sources were handled (inside and outside the glovebox lines); 

• The defect analysis glovebox area where final units were disassembled; 

• The storage vault where heat sources, partially assembled units, and completed units awaiting 

shipment were stored; 

• The exhaust plenum [with high-efficiency particulate air (HEPA) filter] and ventilation exhaust 

stacks. 

The filters for the receiving room, defect analysis box, and the storage vault were changed on 

Mondays, Wednesdays, and Fridays, and the air rotometers were checked at these times. Monthly 

changes occurred for HEPA filters and gloveboxes, and the air rotometers were also checked at this 

time. The air monitors were set at 10- or 15-cpm alarm points. They were checked with 239Pu check 

sources on a monthly basis. 

The Eberline alpha continuous air monitors were set at a fraction of the air concentration limits. They 

operated to monitor specifically for 238Pu or 239Pu and had background subtraction and alarming 

capabilities. Since the 1970s, Pinellas employed high-technology silicon barrier detector systems 

connected to a single-pulse analyzer, which enabled the monitoring of air for potential releases of 

plutonium (EIC 1970). The Pinellas Plant continued to upgrade the plutonium air monitoring systems 

over the years (Dixon 1988a,b). 

Air sampling occurred at nine points (probably identical to the real-time air-sampling points listed 

above). These samplers operated continuously and were changed monthly. Following 3 days of 

decay, they were counted for gross alpha and composited quarterly for 238Pu analysis. In the 1988 

plutonium bioassay report, it was indicated that no plutonium had been found detected in the air, on 

heat sources, or on completed RTG units (Weaver 1989). 
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 16 of 39

5.3.3

Tritium and plutonium were the only radioactive materials that bioassay monitoring was routinely 

performed for. Nonroutine bioassay monitoring might have been performed for other radioactive 

materials; however, no documentation regarding other potential types of bioassay monitoring has 

been found. 

Bioassay Monitoring Practices

5.3.3.1 Records Interpretation

The only interpretation issue that is relatively common in the bioassay records has to do with the 

continuation sheets for the tritium urinalysis results, which are used when the number of urine 

samples for a given year exceed the allotted space on the primary datasheet. These continuation 

sheets sometimes do not include any information regarding the year that the bioassay results were 

collected. In those instances, the dates that are provided only include the month and day that each 

urine sample was collected. Fortunately, the continuation sheets appear to always follow the sheet 

with the data from the first part of the year. This can also be confirmed by comparing the end dates of 

the first sheet with the start dates of the second sheet. The start dates of the second sheet should 

pick up where the end dates of the first sheet left off.

5.3.3.2 Tritium Bioassay Monitoring

Since about 1986, the bioassay program at Pinellas was based upon ANSI standard N13.14-1983 

(ANSI 1983). Participation was determined based on the recommendations in the ANSI standard and 

included:

• Anyone with the potential to receive 100 mrem/year from tritium;

• Declared pregnant workers likely to receive more than 50 mrem/gestation period;

• Minors, visitors, and members of the public likely to receive more than 50 mrem/year;

• All personnel who worked with or handled tritium-contaminated systems or equipment.

The routine sampling frequency was determined by the extent of possible exposure. In the earlier 

years from about 1957 to 1972, it appears that the frequency was usually weekly, but monthly or daily 

samples could have been taken. In later years, the frequency most likely followed the criteria as 

stated in the Pinellas internal dosimetry TBD (Burkhart 1995a; GE 1984ab, p. 2) as follows:

Daily or on each performance:

• Work on open neutron generator tubes or tube processing equipment;

• Maintenance on vacuum pumps, glove boxes, or exhaust systems including the Tritium 

Recovery system (TRS);

• Instances of area contamination (1984 version);

• Packaging and disposal of radioactive waste (1984 version).

Weekly:

• Operation of contaminated processing or analysis equipment; 

• Decontamination of materials and facilities;

• Packaging and disposal of radioactive wastes;

• Mass spectrometers and tritide films (1984 version).

Monthly:

• Handling of processed tubes (slight potential of measurable exposure).
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 17 of 39

Pinellas required tritium bioassays for contractors and nonroutine work in tritium areas covered by 

work permits at the conclusion of the work. This occurred throughout the work history at the Pinellas 

Plant.

Beginning at least as early as 1966, urine samples to be analyzed for tritium were submitted on a 

daily basis at the end of a worker’s shift, or at the beginning of the next shift for each individual 

worker. Spot samples were composited from each shifts’ individual samples to be counted on the 

next shift. If the samples were weekly, daily samples were composited to give the weekly sample that 

was then sampled and counted. 

A relative biological effectiveness (RBE) of 1.7 was used to calculate tritium exposures for the years 

before 1976. This was suggested by the National Bureau of Standards (NBS), ICRP, and National 

Council on Radiation Protection and Measurements-published references of that time. Calculations 

for 1976 and subsequent years were made using an RBE of 1.0 (Holliday ca. 1976). Exposures prior 

to 1976 were later adjusted to an RBE of 1.0. In the exposure records, most of the tritium dose 

records were adjusted from 1957 to 1975 by dividing by 1.7, as indicated in handwritten calculations 

in the claimants’ DOE dose files. 

Prior to 1987, a 5 mrem/μCi/L infinite dose factor, which was based on an effective half-life for tritium 

in the body of 12 days, was used for the tritium dose calculations (Holliday 1983). In November of 

1986, the infinite dose factor was recalculated as 4.21 mrem/μCi/L, which was based on an effective 

half-life of 10 days for tritium in the body (GE 1986, Weaver 1994). The use of the new infinite dose 

factor appears to have been implemented in 1987, and was used for the tritium dose calculations 

throughout the remainder of the Pinellas Plant’s history (GE 1986, Weaver 1994). Both of the infinite 

dose factors were based on the following equations (GE 1986).

(5-1)

 (5-2)

 (5-3)

where

DCF∞ - daily intake to infinite dose conversion factor [mrem/(μCi/day)]

MBW - mass of total body water in standard man (42,000 g)

QF - quality factor for radiation type (1 rem/rad)

VBW - volume of total body water in standard man (L)

dW - density of water (1 g/ml)

λE - effective removal rate for tritium in the body (day-1)

T1⁄2E - effective half-life of tritium in the body (days)

42L

d

mL

L x1M 10

V

W

3

BW

BW =





 





=

−

( )

E

2

1

E T

ln 2 =λ





 



 λ





 









 





















































μ =

− −

∞

rad

erg g/ M 100QF

rem

mrem 1000

Mev

erg x6.1 10

disintegration

Mev x7.5 10

day

min 1440

Ci

dpm .2V 22 x 10

DCF

BW E

6 3 6

BW
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 18 of 39

For T1⁄2E = 12 days: λE = 0.05776 day-1 and DCF∞ = 5.048 ≈ 5 [mrem/(μCi/day)] 

For T1⁄2E = 10 days: λE = 0.06931 day-1 and DCF∞ = 4.206 ≈ 4.21 [mrem/(μCi/day)] 

Since 1986, the method for calculating internal tritium doses at the Pinellas Plant was based on ANSI 

Standard N13.14-1983 (ANSI 1983). Prior to that, the following equation was used to calculate the 

internal tritium doses (Holliday 1983). 

D∞ = DCF∞[(Cie−λE(ΔT−1)] − (Ci−1e−λET0)

(5-4)

where

D∞ - infinite dose (rem)

DCF∞ - daily intake to infinite dose conversion factor [5 mrem/(μCi/day)]

Ci - tritium concentration in most recent urine sample (μCi/L)

Ci-1 - tritium concentration in prior urine sample (μCi/L)

ΔT - elapsed time between the sample collection dates (days)

T0 - time after prior sample (1 day)

T1⁄2E - effective half-life of tritium in the body (12 days)

λE - effective removal rate for tritium in the body (0.05776 day-1)

The GEDOSE computer program was written in 1988 to process both external and internal dosimetry 

data and doses. The computer program’s trigger for performing internal dose calculations was 0.1 

μCi/L. If this level was exceeded by any urine sample submitted during a calendar year, an internal 

dose was calculated for the worker using all of that year’s sample results. If none of a worker’s urine 

sample results for a given year exceeded the trigger level, an internal dose of zero was assigned for 

that year (Burkhart 1995a).

5.3.3.3 Plutonium Bioassay Monitoring

The Safety Analysis Report for the Radioisotopic Thermoelectric Generator (RTG) Facility indicates 

that personnel assigned to work in the east portion of Building 400 (i.e., the RTG operations portion) 

were required to submit a preoperational urine sample and annual samples thereafter, because of the 

presence of the encapsulated plutonium sources at the site (GE1982a, Holiday 1983). The available 

plutonium bioassay data also confirms that this was being done.

For urine samples being analyzed for plutonium, the reported sample size was what was received. If 

the sample received was less than 800 mL, the entire sample was analyzed. If more than 800 mL 

was received, only an 800 mL aliquot was analyzed. 

The criteria used by the Pinellas Plant to evaluate the plutonium bioassay results indicated that no 

plutonium intakes occurred at the site. In addition, the Pinellas Plant never assigned any internal 

plutonium doses based on its assessment of the plutonium bioassay data. 
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 19 of 39

5.4 IN VITRO BIOASSAY DETECTION LEVELS AND ANALYSIS METHODS

Urine sampling was the only in vitro bioassay method employed at the Pinellas Plant.

5.4.1

5.4.1.1 Tritium Urinalysis Detection Levels

Urinalysis Detection Levels

Tritium urinalysis is only capable of detecting intakes of soluble forms of tritium (i.e., tritium gas, 

tritiated water (HTO), uranium tritide, and certain organically bound tritium compounds [e.g., methane, 

acetone, and octane]). Because urinalysis is ineffective for detecting insoluble forms of tritium (DOE 

2006), potential intakes of insoluble forms of tritium are considered to be unmonitored at the Pinellas 

Plant. 

Table 5-3 shows that the tritium urinalysis MDCs and reporting levels varied over the years at the 

Pinellas Plant. When available, the information in an individual worker’s dosimetry records is to be 

used. If claim specific details regarding the urine sample MDCs and reporting levels are not available, 

the dose reconstructors should use the higher of the two values in Table 5-3 for a given period as the 

default MDC for the internal dose calculations. For example, a default MDC of 0.90 μCi/L would be 

used for 1957, and a default MDC of 0.67 μCi/L would be used for 1963. Note that using reporting 

levels that are below the MDC could result in an underestimate of the missed internal dose. 

Table 5-3. Default tritium urinalysis MDCs and reporting levels.

a,b

Period

MDC

μCi/L

Reporting

levelc

μCi/L 

1957–Jul 1958 0.90d 0.3

Aug 1958–Dec 1958 0.90d 1.0

Jan 1959 0.80d 1.0

Feb 1959–Dec 1959 0.80d N/A

1960–Jun 1961 0.50d N/A

Jul 1961–1968 0.50d “<”e

1969–1970 0.50f 0.67

1971–Mar 1974 0.20d 0.67

Apr 1974–Dec 1974 0.20d 0.10

1975–1986 0.10 0.10

1987–1989 0.01 N/A

1990–1997 0.006g N/A

a. Sources: Claims; GE 1968, 1971, undated a, undated b, undated c; Holliday 

1983; Burkhart and Richardson 1986; Burkhart 1995a.

b. Urinalysis method was liquid scintillation counting for tritium throughout 

Pinellas Plant site history.

c. All reporting levels are based on what was observed in the available 

dosimetry records for the Pinellas Plant, and appear to have been 

consistently used for all workers that were monitored for tritium intakes; N/A 

– not applicable, because no reporting level was used during this period. 

d. MDC value calculated by ORAUT from data in urine sample counting logs

and procedures. 

e. If no specific reporting level value can be determined from the dosimetry 

records assume the reporting level for this period was <0.67 μCi/L.

f. Assumes MDCs were not improved after 1968.

g. MDC value obtained from Burkhart 1995a.
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 20 of 39

5.4.1.2 Plutonium Urinalysis Detection Levels

Plutonium urinalysis detection levels varied for individual sample results and were reported in terms of 

MDCs. With the exception of the 1975 and 1976 plutonium bioassay data, the MDCs associated with 

the specific sample results are always reported with the plutonium bioassay results. However, to be 

favorable to the claimant, the best available MDC will be used to evaluate the Pinellas Plant’s isotopic 

plutonium urinalysis data. As a result, the isotopic plutonium MDCs for the Los Alamos National 

Laboratory (LANL) will be used for evaluating the plutonium urinalysis data against the criteria 

specified in Section 5.7.3.

The LANL was one of the forerunners in developing plutonium bioassay methods and was one of the 

few sites performing an isotopic plutonium analysis on urine samples when the Pinellas Plant began 

using this analysis method. As a result, the analysis method that was used at the Pinellas Plant was 

likely based on the method being used at LANL, which makes the LANL MDCs the best available 

MDC information available for use with the Pinellas Plant data. The TBD for the Los Alamos National 

Laboratory – Occupational Internal Dose indicates that the MDC for both 238Pu and 239Pu is 0.03 

pCi/24-hr sample for the 1975–1990 timeframe (ORAUT 2009). An MDC of 0.03 pCi/24-hr sample is 

equivalent to 4.8 x 10-5 dpm/ml and 2.1 x 10-11 μCi/ml.

5.4.2

5.4.2.1 Analysis Methods for Tritium

Analysis Methods for In Vitro Urine Samples

Liquid scintillation counting (LSC) was used to analyze the urine samples for tritium. A sample 

consisted of 0.2 mL of urine combined with 16.5 mL of aqueous scintillation mix (1 L toluene, 0.2 L 

ethanol, 3.5 g PPO, 0.12 g POPOP) (GE 1964). From 1957 through at least 1964, urine samples 

were typically counted for 5 minutes (GE 1957, 1958, 1959, 1960, 1964). By 1966, the typical count 

times for the urine sample had been reduced to 1 minute (GE 1966, 1968, undated a). Counting 

efficiencies for the years of 1957–1960 ranged from about 5% up to about 13% (GE 1957, 1958, 

1959, 1960). By 1964, counting efficiencies ranged from about 13% up to about 18% (GE 1964, 

1968, 1971).

Beginning at least as early as 1966, daily urine samples were composited for screening purposes. 

Aliquots of the daily urine samples collected at the end of each shift from individual workers were 

composited with aliquots of the daily urine samples from up to 40 workers. These composite samples 

were then analyzed during the following shift. If a composite sample was above the reporting level, 

the composite group was subdivided into 4 groups and 4 new composite samples were created and 

analyzed. For the subdivided composite samples that were above the reporting level, the individual 

workers’ urine samples used to make the subdivided composite sample were analyzed individually. 

Weekly and monthly urine samples were not composited with urine samples from other workers (GE 

undated a). 

A review of the available bioassay records indicates that very few daily urine sample results were 

reported, even for workers that were likely on a daily sampling frequency. The data indicates that the 

daily samples for individual workers may have been composited into a weekly composite sample or 

that only one of the urine samples collected during a given week was analyzed individually. The 

available procedures discuss the analysis of the weekly sample for individual workers that were on a 

daily urine sampling frequency, but no specifics about the weekly sample could be found. A few daily 

sample results do show up in the records periodically, but they could be the result of the analysis of 

the individual samples that would have been performed after a composited screen sample was found 

to be over the reporting level. 
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 21 of 39

The composite samples were prepared by combining a 1 mL aliquot from each worker’s daily urine 

sample in a paper cup. A procedure dating from the 1970s was first to add charcoal to the solution to 

decolorize the solution and filter the composite through Whatman #1 filter paper. Then 0.2 mL of the 

composite sample was pipetted into a vial containing 16.5 mL of the aqueous scintillation mix. The 

composite sample vials were then placed in the freezer for cooling and counted for 10 minutes on a 

Beckman multichannel liquid scintillation counter (GE undated a). 

The 1970’s urine samples for individual workers were processed in a similar manner with the following 

exceptions: 1) the samples were not composited, 2) the charcoal was added directly to the individual 

sample container, 3) 5 mL of the urine sample was poured into a paper cup, and 4) the samples were 

initially counted for 1–2 minutes. On the Beckman multichannel liquid scintillation counter, the preset 

error control for channel 1 was set to 15%, the preset time was set to 1–2 minutes, and the action 

control switch was set to cycle repeat. When the initial count of an individual’s urine sample result 

was above the reporting level, the result was verified by processing and counting a second sample for 

5 minutes (preset error ± 5%) (GE undated a).

5.4.2.2 Analysis Methods for Plutonium

Urine samples being analyzed for plutonium were prepared in a number of steps. The first step was a 

1-2 week dissolution process. The second step was a solution purification with the use of an anion 

exchange resin column. The third step was a 1-day electroplating. The final step was alpha 

spectroscopy. In addition, Pinellas used National Institute of Standards and Technology (NIST)-

traceable sources of 238Pu and 239Pu to set up the ROIs (Christy 1988).

In 1988, two major changes to plutonium sampling and analysis were made. The first change was 

from a one-channel analysis to an energy region of interest (ROI) technique (Weaver 1989a). The 

use of polished stainless-steel planchets was implemented (Christy 1988). 

By 1989, the Pinellas chemistry counting laboratory used a Series 100 Canberra alpha spectroscopy 

system [as shown in a sample spectrum comparison to NIST 238Pu and 239Pu source spectrums (GE 

1989b)]. This alpha spectroscopy system likely used a silicon barrier detector system. The energy 

ROI for 238Pu was 5,409.7 to 5,535.5 keV and that for 239Pu was 5,063.4 to 5,186.6 keV. The 1988 

and 1989 bioassay result folders (GE 1987–1990) show that the ROIs varied. The 200-keV interval 

between the ROIs for 238

Pu and 239Pu indicates that very good resolution was attained for this 

counting system. A 24- to 72-hr counting time was usually used, but in 1989, a 24-hr counting time 

was used. A bioassay recount was made for 6 days or 144 hr to improve counting statistics (Weaver 

1989b). 

Pinellas conducted a quality assurance check on the plutonium bioassay program by comparing 

several samples of blind 238Pu/239Pu artificial urine samples with the results from the DOE Mound 

Laboratory. These comparisons were conducted from at least 1987 though 1990. The results 

compared favorably (Burkhart 1990b). 

Plutonium concentrations in urine for 238Pu and 239Pu were calculated with the following formula (GE 

1987–1990 pp. 83–93):
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 22 of 39

(Nnx)(C242)(V242)(1 x 10-6 μCi/pCi)

Cx = (Nn242)(Vs)(2.22 dpm/pCi) (5-6) 

where

x - atomic mass number for the applicable Pu isotope (i.e., xPu = 238Pu or 239Pu)

Cx - concentration of xPu in the urine (μCi/mL)

Nnx - net counts accumulated during a count time of Tg for the xPu region of 

interest

C242 - concentration of 242Pu in tracer (dpm/mL) 

V242 - quantity of tracer that sample was spiked with (mL)

Nn242 - net counts accumulated during a count time of Tg for the 242Pu region of 

interest

Vs - volume of urine used for sample (mL)

Tg - gross count time for the measurement

The uncertainties associated with the 238Pu and 239Pu in urine results were based on the formulation in 

Equation 5-7 (GE 1987–1990 pp. 83–93). Based on Equation 5-7, the uncertainty was calculated at 

the 95.44% confidence level (i.e., at the 2σ confidence interval). 

 2(Nnx)

1⁄2(C242)(V242)(1 x 10-6 μCi/pCi)

 Ux = (Nn242)(Vs)(2.22 dpm/pCi) (5-7) 

where

x - atomic mass number for the applicable Pu isotope (i.e., xPu = 238Pu or 239Pu)

Ux - uncertainty of the xPu in urine result (± μCi/mL)

Nnx - net counts accumulated during a count time of Tg for the xPu region of 

interest

C242 - concentration of 242Pu in tracer (dpm/mL) 

V242 - quantity of tracer that sample was spiked with (mL)

Nn242 - net counts accumulated during a count time of Tg for the 242Pu region of 

interest

Vs - volume of urine used for sample (mL)

Tg - gross count time for the measurement

5.5 IN VIVO MINIMUM DETECTABLE ACTIVITIES AND ANALYSIS METHODS

There was no in vivo monitoring at the Pinellas Plant.

5.6 INTERFERENCES AND UNCERTAINTIES

5.6.1

Uncertainties or errors for tritium bioassay measurements were usually not stated in the personnel 

records or database, except for 1972-1980 at a level of ±10% error. MDCs were calculated based 

Tritium Bioassay
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 23 of 39

upon a 95% confidence level and samples were counted to between a 10% to 15% error as indicated 

in Pinellas Plant procedures (GE undated). 

Cross-contamination of the tritium urine samples was a potential interference that would have resulted 

in an overestimate of a worker’s potential soluble tritium exposure. In about 1985, it was stated that 

one positive tritium urine sample followed by a negative sample would negate the first positive sample 

if given in the same day at the Pinellas Plant. If the urine sample for tritium was unattainable the 

same day, the employee was asked to give the sample within the first hour of being on site the next 

day. It is likely that most personnel did this, thus minimizing the possible cross-contamination issue 

for tritium bioassay samples. 

5.6.2

For plutonium bioassay, uncertainties were typically reported with the plutonium urinalysis results after 

1976. 

Plutonium Bioassay

There should have been no issue of cross contamination with plutonium samples because they were 

taken annually and not likely after recent work with the plutonium sources. However, plutonium from 

the atmospheric testing of nuclear weapons is an interference that was encountered with low-level 

plutonium measurements. The interference from weapons related plutonium, which is dominated by 

239/240Pu, is readily discernable from the Pinellas Plant plutonium that was dominated by 238Pu.

5.7 ASSESSMENT OF INTAKES AND DOSES

The Pinellas Plant had an extensive bioassay program from the beginning of operations. Urinalysis 

started in 1957. Although the earlier techniques had their sensitivity limitations, the detection 

sensitivity seemed to keep pace with the fast-paced regulatory and safety changes. Seldom did 

workers achieve or surpass the site action levels of the radionuclides of concern. All Pinellas Plant 

workers that were potentially exposed to tritium and plutonium were likely monitored for potential 

internal dose, with the only likely exception being maintenance workers. Information provided by 

employees at the Pinellas Plant indicates that maintenance workers were often not monitored for 

internal dose when working in areas with tritium. 

5.7.1

5.7.1.1 Assessment of Soluble Tritium Exposures

Workers Monitored for Soluble Tritium Exposure

For the periods that a worker was monitored for soluble tritium exposures, the potential exposures 

should be assessed in accordance with the recommendations in ORAUT-OTIB-0011 (ORAUT 2004) 

and ORAUT-OTIB-0060 (ORAUT 2007) using the worker’s urine sample data. 

5.7.1.2 Assessment of Potential Unmonitored Exposures to Insoluble Tritium

At the Pinellas Plant, insoluble tritium (i.e., certain metal tritides and certain organically bound tritium 

compounds) were handled only in areas where the more dispersible and more soluble forms of tritium 

[e.g., elemental tritium (HT), tritiated water (HTO), etc.] were also present. Given that a review of the 

available dosimetry records indicates that the Pinellas Plant routinely monitored workers with any 

potential for soluble tritium exposures, any workers with potential metal tritide exposures would have 

been monitored for soluble tritium exposures. Therefore, insoluble tritium exposures at the Pinellas 

Plant are only assessed for the periods that workers were monitored for soluble tritium exposures. 

Given that the least frequent routine monitoring frequency for soluble tritium was monthly, the 

minimum period for assessing insoluble tritium exposures for intermittently monitored workers should 

be 1 month.
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 24 of 39

The 1957 through 1973 monthly health physics reports (GE 1957–1973) were reviewed to determine 

tritium contamination controls at the Pinellas Plant. These reports indicate that as early as 1959, 

areas greater than 2 × 10-5 μCi/in2 (688 dpm/100 cm2

) were recommended for decontamination. In 

1969, the control limit was reported as 440 dpm/100 cm2

. This indicates that a routine contamination 

control program was in place throughout the history of the site and that it would be unlikely to see high 

contamination levels for extended periods.

The monthly health physics reports also provide information on the maximum tritium surface 

contamination levels. Between 1957 and 1973, the highest surface contamination level reported was 

in 1970, 4.4 × 106 dpm/100 cm2 (10,000 times the control limit). The next highest value was reported 

in 1959, 1.4 × 106 dpm/100 cm2 (3.3 × 10-2 μCi/in2

). The majority of the rest of the annual maximum 

surface contamination levels reported are at least an order of magnitude lower than these two. 

However, the following assumptions are made to bound the potential unmonitored metal tritide 

exposure: 

• Constant tritium surface contamination level of 4.4 × 106 dpm/100 cm2 (4.4 × 108 dpm/m2

) 

(GE 1957–1973);

• Resuspension factor of 1 × 10-6/m (ORAUT 2008). 

Based on these assumptions, a constant tritium air concentration of 440 dpm/m3 was possible in the 

areas where tritium was handled. Assuming a breathing rate of 1.2 m3

/hour and the exposure time 

assumption of 2,600 hours (based on a review of telephone interviews provided by former workers, 

50-hour weeks were routine), annual inhalation and ingestion intake rates for insoluble tritium were 

calculated and are provided in Table 5-4. The values provided in Table 5-4 should be assigned to 

account for potential insoluble tritium intakes for the periods that a worker was exposed to tritium in 

addition to any soluble tritium intakes that were assessed based on tritium bioassay data. For partial 

years of employment or monitoring, the daily intake values in Table 5-4 can be used. 

Table 5-4. Insoluble tritium intake rates. 

Hours worked 

per yeara

Annual intake rates (pCi/yr) Daily intake ratesb (pCi/d)

Inhalation Ingestionc Inhalation Ingestionc

2,600 6.18E+05 1.29E+04 1.69E+03 3.53E+01

a. Based on a 10-hour workday, 5 days a week, for 52 weeks a year.

b. These daily intake rates are calculated for a calendar year (i.e., 365 d/yr) and can be 

entered directly in the IMBA computer program without any conversions.

c. Based on the recommendation in OCAS-TIB-0009 (NIOSH 2004) for best estimates.

The available information indicates that insoluble tritium compounds representing lung absorption 

Types M and S were present at the Pinellas Plant. Because there is insufficient information to 

indicate which insoluble tritium compounds a worker might have been exposed to, potential exposures 

should be assessed for both types of insoluble tritium compounds.

5.7.2

Workers at the Pinellas Plant were likely only exposed to on-site levels of environmental radioactivity

during the unmonitored periods of their employment. 

Workers Not Monitored for Soluble Tritium Exposure

5.7.3

Because only encapsulated plutonium sources were used at the Pinellas Plant and because there is 

no indication of any plutonium releases at the site, plutonium intakes were very unlikely at the Pinellas 

Workers Monitored for Plutonium Exposure
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 25 of 39

Plant, and were most likely only limited to the personnel performing the receipt inspections on the 

plutonium sources. Therefore, no missed plutonium doses need to be assigned for the monitored 

Pinellas Plant workers when all of their plutonium bioassay results were negative. In the event that a 

case is encountered where the Pinellas Plant worker had any

The following are criteria to confirm that a Pinellas Plant plutonium bioassay result is positive: 

positive bioassay results, as defined by 

the criteria below, potential plutonium intakes and doses will need to be assessed and assigned in 

accordance with the recommendations of ORAUT-OTIB-0060 (ORAUT 2007). 

1. The plutonium bioassay result is greater than the MDC;

2. The plutonium result is not from a preoperational/baseline sample, because 

preoperational/baseline bioassay measurements are not an indication of exposure; 

3. The 238Pu result is greater than the 239Pu result;

4. The sample result is greater than its reported 2σ uncertainty value, such that the lower part of 

the range does not include zero. 

Because uncertainty values are not available for the 1975 and 1976 plutonium bioassay results, the 

1975–1976 plutonium bioassay results are considered to be positive if they meet the first three criteria 

above. Even though some of the documentation on the Pinellas Plant’s internal dosimetry practices 

indicates that the plutonium bioassay results were being calculated at the 95% confidence level (i.e., 

at a 1.96σ uncertainty), the reported uncertainties were actually calculated at the 95.44% confidence 

level (i.e., at a 2σ uncertainty) (GE 1975–1986, 1987–1990, Holliday 1983). The equation used for 

reported MDLs (Equation 5-5 above) indicates that the MDLs were calculated to the 95% confidence 

level.

If the plutonium bioassay results provided by the DOE are not in units of concentration and do not 

include the sample volumes analyzed, the dose reconstructor will need to obtain the concentration 

data from captured Pinellas Plant records. For the years of 1975–1990, compilations of all of the 

Pinellas Plant’s plutonium bioassay results are available in two captured documents (GE 1975–1986, 

1987–1990) with the exception of the 1984 data. At the time of this revision, no similar compilation of 

plutonium bioassay data has been found for 1984. Note that an automated search of the Site 

Research Database records may have already located the records for a specific claim and provided 

them as Personnel Exposure files under the DOE files in the NIOSH OCAS Claims Tracking System 

(NOCTS).

5.7.4

As indicated in Sections 5.2.2, 5.3.3.2, and 5.7.3, plutonium intakes were very unlikely at the Pinellas 

Plant. Therefore, no potential plutonium exposures need to be assessed for Pinellas Plant workers 

that were not monitored for plutonium exposures. 

Workers Not Monitored for Plutonium Exposure

5.8 INCIDENT HISTORY

Table 5-5 lists a chronology of some of the unusual events that occurred and the resultant radiological 

releases (if known) from startup to 1989. Most of the releases were through the environmental stack 

release system, but some elevated tritium levels occurred in other areas. 
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 26 of 39

Table 5-5. Chronology of unusual events and significant activities in relation to internal dosimetry. 

Date Description

Curies of 

tritium 

released

12/57 Foot monitor installed in Area 108

12/10/57 Operator error in reading manometer in Room 18 458

02/11/58 Error estimating amount of tritium remaining in charging system in Room 18 1,253

07/08/58 Glass system breakage in Room 22 280

03/07/58 Glass system breakage in Room 18 567

08/16/58 Operator error with tritium loader valve position – Room 21 780

08/18/58 Glass manifold breakage – Room 21 1,180

02/10/59 Operator error in valve positioning – Room 8 286

02/20/59 Hand contamination – operator not wearing gloves

02/21/59 Area contamination – operator broke glass system

03/12/59 Operator contaminated during system cleaning by another worker (Room 14)

06/04/59 Personnel error working on SECS test system – Room 21 753

06/05/59 Area contamination-diffusion pump exploded in hood 14

06/18/59 Excess air released from tritium loading system 423

01/1960 Operator error – stopcock left open on tritium loading system 40

02/05/60 Glass manifold broke from strain 72

02/11/60 Operator error – stopcock left open on tritium loading system 308

03/25/60 Operator error caused exposure to 3 employees

05/14/60 Broken flask caused area contamination – Room 10

06/21/60 Ion gauge exploded – room 16

07/08/60 Sample bulb dropped – room 23 6.8

07/13/60 Manifold shattered, exposing worker – room 23

08/12/60 Contamination spread TiH2 in area 108 from broken flask

09/10/60 Requirement for full anti-contamination clothing in Area 108 reduced to lab coats for 

normal production operations

11/14/60 Began using NBS Handbook 69 (NBS 1959) for maximum permissible 

concentrations (MPCs)

04/61 Area contamination from system breakage

10/62 A modified personal monitor was installed in Area 108

12/62 Breathing air supply line connected to Area 108 exhaust duct

09/63 Employees found falsely identifying urine samples

03/30/65 Broken flask – room 9

03/20/65 Flask explosion – room 12

05/66 SECS cold water removal problems 252

01/27/67 Glove box vacuum pump oil degassed 32

10/12/67 Personnel contamination – O-ring mishandled – room 18

06/18/68 Acid cleaning explosion – Area 181

02/69 Leaking flange at absorption pump in Area 108 8 

02/69 Area contamination when pump exhaust lines were cut during hood removal – room 

2 

8 

11/05/69 Area contamination in building 400 associated with D-bed ? 

11/69 Building 400 cell #3 contaminated with TiH2 from used flask storage; all flasks 

moved to burning pad west of 400.

01/70 Area contamination/personnel exposure from flaking tube part in gas lab

02/70 Area contamination from pressurized sorb pump (air expansion) – room 2
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 27 of 39

Date Description

Curies of 

tritium 

released

11/20/70 Area contaminated when operator used vacuum cleaner on ScH2 dust in Area 182D

12/28/70 SECS column saturated due to air leak in area 108 117

03/12/71 Copper gasket uncovered in Room 18 hood – high internal dose 7.3

10/21/71 Tritium release from improperly baked evaporator system in area 182D 129

11/10/71 Area contamination from T loaded disc –Auger spectrometer sample – Area 154 0 

12/01/71 High internal exposure – room 18 hood work 1 

04/72 Area contaminated from liquid discharge in area 182D 1.5

08/03/72 Leaking absorption pump 12

05/73 Area contaminated with ErH2 film in Areas 156, 157, and 158

11/73 Fire in boom box- building 200 0 

01/31/75 Improper valve closure on uranium bed 150

02/10/75 Absorption pump leak – area 182D 42

01/30/76 Contaminated 6 inch valve 0 

04/13/76 Oven fan blade broke tubes 0 

02/77 Packaging fixtures in area 182D glove box 28

09/11/79 Work in Room 18 hood – internal dose 5.7

04/80 Area contaminated from film flaking – Area 158B 0 

08/80 Contaminated electron microscope 0 

04/81 Three waste drums found out gassing during truck loading were removed to Area 

108; a procedure to check all drums for out gassing prior to transfer to the storage 

building was set up. 

02/25/82 Tritium recovery system (TRS) valve left in wrong position after maintenance 8.6

04/20/82 Operator left TRS valve in wrong position after maintenance – area 108 48

05/24/82 TRS valve left in wrong position after maintenance – area 108, exhaust unit 513 9.5

09/01/82 Sample bulb leak 3 

01/05/83 Tritium storage bed oxidation problems 130

01/19/83 Absorption pump leak – area 208 9 

04/05/83 Bed heater control failure – Area 108 0 

05/84 Tritium air monitor system in Area 182 was rearranged

07/25/84 Absorption pump sieve dumped into drum in area 108 67

12/09/85 Sorb pump overheat – area contaminated 0 

06/24/86 Mass spectrometer oil change – workers exposed to tritium gas 1.5

11/04/87 Test of oxygen regeneration unit 12

02/11/88 Leaking sample bulb in area 108 8 

05/05/88 Purge left on over third shift in Area 108 – SECS overpressure 2.7

09/88 Lab area release over 2-week period 16.2

01/06/89 Water in SECS line vented in area 182D 1 

09/07/89 Loss of control of radioactive material 0 

Several examples of incidents are described to provide perspective on the operational health physics 

safety responses. On November 5, 1969, Cell #3 in Building 400 was found to be contaminated with 

HTO from stored leaking uranium beds. Some floors were found to have 4,000 dpm/100 cm2 of HTO 

contamination, but the air concentration of the exhaust measured about 600 μCi/m3

. Bioassays of 

personnel in Building 400 revealed a maximum level of 5.3 μCi/L and a maximum infinite personnel 

internal dose of 50 mrem (GE 1969). 
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 28 of 39

On September 11, 1979, a release of 5.6 Ci of HTO from the main exhaust stack occurred at 

approximately 11:43 a.m. The release resulted from nonroutine modifications being made to the 

uranium bed process system in Hoodroom 18, Area 108. The technician performing the work 

received a forearm skin exposure that resulted in a body burden of 62.2 μCi/L. This was calculated to 

236 mrem. The reasons for the incident included inadequate system modification procedures, a 

nonfunctional hood monitor, inadequate testing of the hood monitor, and improper use of personal 

protective equipment. An operator alert system for the HTO monitor was added to the monitoring 

equipment at about this time in response to this incident. The identification of the problem occurred 

due to the stack monitor alarm and subsequent security notification to health physicists, who traced 

the release to Room 108 (Mauer 1979).

On about March 22, 1995, leak cylinders containing hydrogen, Freon, and argon cylinders (which had 

been sent to an outside vendor) were found contaminated with HTO to a level of about 

600,000 dpm/100 cm2

. The vendor was notified, and Pinellas conducted a contamination survey at 

the vendor’s site and provided urinalysis for all requested personnel (Burkhart 1995b). 

Each of the above incidents identifies that Pinellas took some routine operational actions to maintain 

exposures below the applicable limits of the time. Routine air monitoring and contamination 

monitoring usually led to identification of problems. It is evident that if personnel were suspected of 

internal exposure to tritium, they were promptly asked to submit a urine sample. Work orders quite 

often required bioassays after completion of the work (GE 1992, 1978–1989). In addition, outside 

contractors were given tritium bioassays from the start of Pinellas operations in 1957.

5.9 ATTRIBUTIONS AND ANNOTATIONS

All information requiring identification was addressed via references integrated into the reference 

section of this document.
Document No. ORAUT-TKBS-0029-5 Revision No. 01 Effective Date: 04/01/2011 Page 29 of 39

REFERENCES

ANSI (American National Standards Institute), 1983, American National Standard for Dosimetry –

Internal Dosimetry Programs for Tritium Exposure – Minimum Requirements, N13.14-1983, 

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Author unknown, 1973, Radioanalyses Sensitivities, November 8. [SRDB Ref ID: 12833, p. 24]

Author unknown, 1987, Response to Kanne Air Monitor Alarms, November 4. [SRDB Ref ID: 13074]

Author unknown, undated a, Item #7: The date plutonium was first delivered on-site and the nature 

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GE (General Electric Company), 1964, collection of Weekly Urinalysis Radiation Survey, Pinellas 

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GE (General Electric Company), 1965, collection of “Individual Radiation Exposure Record,” Pinellas 

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GLOSSARY

activity median aerodynamic diameter (AMAD)

The diameter of a unit density sphere with the same terminal settling velocity in air as that of 

the aerosol particle whose activity is the median for the entire aerosol.

acute exposure 

Radiation exposure to the body delivered in a short period. See chronic exposure. 

alpha radiation 

Positively charged particle emitted from the nuclei of some radioactive elements. An alpha 

particle consists of two neutrons and two protons (a helium nucleus) and has an electrostatic 

charge of +2. 

beta particle

See beta radiation. 

beta radiation 

Charged particle emitted from some radioactive elements with a mass equal to 1/1,837 

that of a proton. A negatively charged beta particle is identical to an electron. A 

positively charged beta particle is a positron.

bioassay

Measurement of amount or concentration of radioactive material either in the body or in 

biological material excreted or removed from the body. Another word for radiobioassay. 

bioassay procedure

A procedure used to determine the kind, quantity, location, and retention of radionuclides in 

the body by direct (in vivo) measurements or by in vitro analysis of material excreted or 

removed from the body.

body burden

The quantity of radioactive material contained in the individual's body at a particular point in 

time.

chronic

Pertaining to low-level intakes received on a continuous basis.

curie (Ci) 

Traditional unit of radioactivity equal to 37 billion (3.7 x 1010) becquerels, which is 

approximately equal to the activity of 1 gram of pure 226Ra.

derived investigation level (DIL)

A value based on the regulatory control level and the minimum detectable concentration of the 

assay method.

dose 

In general, the specific amount of energy from ionizing radiation that is absorbed per unit of 

mass. Effective and equivalent doses are in units of rem or sievert; other types of dose are in 

units of roentgens, rads, reps, or grays
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dose equivalent (H)

In units of rem or sievert, product of absorbed dose in tissue multiplied by a weighting factor 

and sometimes by other modifying factors to account for the potential for a biological effect 

from the absorbed dose. See dose. 

exposure

(1) In general, the act of being exposed to ionizing radiation. (2) Measure of the ionization 

produced by X- and gamma-ray photons in air in units of roentgens.

gamma radiation 

Electromagnetic radiation (photons) of short wavelength and high energy (10 kiloelectron-volts 

to 9 megaelectron-volts) that originates in atomic nuclei and accompanies many nuclear 

reactions (e.g., fission, radioactive decay, and neutron capture). Gamma photons are 

identical to X-ray photons of high energy; the difference is that X-rays do not originate in the 

nucleus. 

insoluble tritium

Less soluble forms of tritium, which have type M or S lung absorption properties (DOE 2006).

intake

The amount of radionuclide taken into the body by inhalation, absorption through intact skin, 

injection, ingestion, or through wounds. Depending on the radionuclide involved, intakes may 

be reported in units of mass, activity, or potential alpha energy.

internal dose or exposure

The dose equivalent received from radioactive material taken into the body (i.e., internal 

sources).

internal dose assessment

An assessment of the intake and associated internal radiation dose to workers based on 

measurements taken in the work environment or from individual bioassay measurements.

In vitro measurement

Measurements to determine the presence of or to estimate the amount of radioactive material 

in the excreta or in other biological materials removed from the body.

In vivo measurement

The measurement of radioactive material in the human body utilizing instrumentation that 

detects radiation emitted from the radioactive material in the body.

ionizing radiation

Radiation of high enough energy to remove an electron from a struck atom and leave behind a 

positively charged ion. High enough doses of ionizing radiation can cause cellular damage. 

Ionizing particles include alpha particles, beta particles, gamma rays, X-rays, neutrons, 

high-speed electrons, high-speed protons, photoelectrons, Compton electrons, 

positron/negatron pairs from photon radiation, and scattered nuclei from fast neutrons. 

isotope 

One of two or more atoms of a particular element that have the same number of protons 

(atomic number) but different numbers of neutrons in their nuclei (e.g., 234U, 235U, and 238U). 

Isotopes have very nearly the same chemical properties.
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kilo-electron volt (keV)

Unit of particle energy equal to 1,000 (1 × 103

) electron-volts.

limit of detection (LOD) 

Minimum level at which a particular device can detect and quantify exposure or radiation. Also 

called lower limit of detection and detection limit or level.

lung absorption type (F, M or S)

Categories for materials according to their rate of absorption from the respiratory tract to the 

blood, which replaced the earlier inhalation clearance classes. Defined by the International 

Commission on Radiological Protection, the absorption types are F: deposited materials that 

are readily absorbed into blood from the respiratory tract (fast solubilization), M: deposited 

materials that have intermediate rates of absorption into blood from the respiratory tract 

(moderate rate of solubilization), and S: deposited materials that are relatively insoluble in the 

respiratory tract (slow solubilization). 

megaelectron-volt (MeV) 

Unit of particle energy equal to 1 million (1 × 106

) electron-volts. 

metal tritide

Metals that absorb tritium atoms in the crystalline structure of the metal. Metal hydrides and 

tritides are the most compact way to store hydrogen or tritium. Because of that, metal tritides 

are most often used as a method of retaining or storing tritium.

minimum detectable amount

The smallest amount (activity or mass) of an analyte in a sample that will be detected with a 

probability, β, of non-detection (Type II error) while accepting a probability, α, of erroneously 

deciding that a positive (non-zero) quantity of analyte is present in an appropriate blank 

sample (Type I error).

minimum detectable concentration (MDC)

The minimum detectable amount expressed in units of concentration.

monitoring

Periodic or continuous determination of the presence or amount of ionizing radiation or 

radioactive contamination in air, surface water, groundwater, soil, sediment, equipment 

surfaces, or personnel (for example, bioassay or alpha scans). In relation to personnel, 

monitoring includes internal and external dosimetry including interpretation of the 

measurements. 

neutron (n) 

Basic nucleic particle that is electrically neutral with mass slightly greater than that of a proton. 

There are neutrons in the nuclei of every atom heavier than normal hydrogen.

neutron radiation 

Radiation that consists of free neutrons unattached to other subatomic particles emitted from a 

decaying radionuclide. Neutron radiation can cause further fission in fissionable material such 

as the chain reactions in nuclear reactors, and nonradioactive nuclides can become 

radioactive by absorbing free neutrons. See neutron. 

occupational dose

An individual's ionizing radiation dose (external and internal) resulting from that individual's 

work assignment. Occupational dose does not include doses received as a medical patient or 
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doses resulting from background radiation or participation as a subject in medical research 

programs.

organically bound tritium (OBT)

A type of tritiated material in which the tritium has formed a chemical bond with an organic 

material, typically via a carbon-tritium bond.

photon

Quantum of electromagnetic energy generally regarded as a discrete particle having zero rest 

mass, no electric charge, and an indefinitely long lifetime. The entire range of electromagnetic 

radiation that extends in frequency from 1023 cycles per second (hertz) to 0 hertz. 

photon radiation 

Electromagnetic radiation that consists of quanta of energy (photons) from radiofrequency 

waves to gamma rays. 

rad

Traditional unit for expressing absorbed radiation dose, which is the amount of energy from 

any type of ionizing radiation deposited in any medium. A dose of 1 rad is equivalent to the 

absorption of 100 ergs per gram (0.01 joules per kilogram) of absorbing tissue. The rad has 

been replaced by the gray in the International System of Units (100 rads = 1 gray). The word 

derives from radiation absorbed dose. 

radiation

Subatomic particles and electromagnetic rays (photons) with kinetic energy that interact with 

matter through various mechanisms that involve energy transfer.

radioactivity

Property possessed by some elements (e.g., uranium) or isotopes (e.g., 14C) of spontaneously 

emitting energetic particles (electrons or alpha particles) by the disintegration of their atomic 

nuclei. 

radioisotopically-powered thermoelectric generator (RTG)

Generator that obtains its power from passive (natural) radioactive decay using thermocouples 

to convert the heat of decay into electricity.

reporting level

A value below which data or results were considered to be too low to record and thus may not 

have been maintained. For example, when the reporting level was “<0.67 μCi/L” the sample 

result was only reported as “<0.67”, or when the reporting level was “0.01 μCi/L” any sample 

result below that value was reported as “0.00” (zero). 

rem

Traditional unit of radiation dose equivalent that indicates the biological damage caused by 

radiation equivalent to that caused by 1 rad of high-penetration X-rays multiplied by a quality 

factor. The sievert is the International System unit; 1 rem equals 0.01 sievert. The word 

derives from roentgen equivalent in man; rem is also the plural.

routine monitoring

Monitoring carried out at regular intervals during normal operations.
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sievert

The special name for the International System unit of dose equivalent. One sievert equals 1 

joule per kilogram, which equals 100 rem.

soluble tritium

All forms of tritium, except for those that have type M or S lung absorption properties.

special monitoring

Monitoring carried out in actual or suspected abnormal conditions (i.e., measurements

performed to estimate the amount of radionuclide deposited in a person when an intake is 

known or is suspected to have occurred). 

spot sample

A single void of urine.

tritium

A radioactive isotope of hydrogen, which has two neutrons. It is also known as or denoted as 

hydrogen-3, 3

H, H-3, and T. 

U.S. Atomic Energy Commission (AEC) 

Federal agency created in 1946 to assume the responsibilities of the Manhattan Engineer 

District (nuclear weapons) and to manage the development, use, and control of nuclear energy 

for military and civilian applications. The U.S. Energy Research and Development 

Administration and the U.S. Nuclear Regulatory Commission assumed separate duties from 

the AEC in 1974. The U.S. Department of Energy succeeded the U.S. Energy Research and 

Development Administration in 1979.

X-ray

See X-ray radiation. 

X-ray radiation 

Electromagnetic radiation (photons) produced by bombardment of atoms by accelerated 

particles. X-rays are produced by various mechanisms including bremsstrahlung and electron 

shell transitions within atoms (characteristic X-rays). Once formed, there is no difference 

between X-rays and gamma rays, but gamma photons originate inside the nucleus of an atom.