Page 1 of 36

DOE Review Release 07/25/2016

Pinellas Plant – Occupational Internal Dose

ORAUT-TKBS-0029-5 Rev. 03

Effective Date: 07/18/2016

Supersedes: Revision 02

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

Document Owner 

Approval: Signature on File

Brian P. Gleckler, Document Owner

Approval Date: 07/13/2016

Concurrence: Signature on File

John M. Byrne, Objective 1 Manager

Concurrence Date: 07/13/2016

Concurrence: Signature on File

Edward F. Maher, Objective 3 Manager

Concurrence Date: 07/13/2016

Concurrence: Vickie S. Short Signature on File for

Kate Kimpan, Project Director

Concurrence Date: 07/13/2016

Approval: Signature on File

James W. Neton, Associate Director for Science

Approval Date: 07/18/2016

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

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

New Total Rewrite Revision Page Change
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 2 of 36

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.

02/21/2012 02 With the exception of Section 5.2.2, all plutonium discussions were 

eliminated from this TBD, discussed at the October 2011 working 

group meeting for the Pinellas Plant Site Profile. Because internal 

exposures to plutonium were unlikely at the Pinellas Plant, the 

plutonium discussions after Section 5.2.2 were determined to be 

unnecessary and a potential source of confusion. 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. 

07/18/2016 03 Sections 5.7.1.1 and 5.7.1.2 were modified to incorporate changes 

agreed upon by the Pinellas Plant Working Group. Additionally, 

Section 5.3.1 was modified to provide more details about the tritium 

contamination smear collection and analysis procedures. Sections 

5.2.1 and 5.3.1 were modified to provide additional details, which 

included changes to the References Section. Section 5.2.1 was 

modified to provide additional details regarding organically bound 

tritium. Section 5.4.2 was modified to provide additional details 

regarding the analysis of the tritium bioassay samples. Incorporates 

formal internal and NIOSH review comments. Training required: As 

determined by the Objective Manager. Initiated by Brian P. Gleckler.

Opti-Fluor® is a registered trademark of General Electric Company or one of its subsidiaries in the 

United States and/or other countries.

Ready-SolvTM is a trademark of Beckman Coulter in the United States and/or other countries.

Tri-Carb® is a registered trademark of PerkinElmer Life Sciences in the United States and/or other 

countries.

Whatman® is a registered trademark of Whatman International Limited in the United States and/or 

other countries.

All other trademarks are the property of their respective owners.
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 3 of 36

TABLE OF CONTENTS

SECTION TITLE PAGE

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

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

5.2 Overview.................................................................................................................................. 8

5.2.1 Purpose........................................................................................................................ 8

5.2.2 Scope........................................................................................................................... 8

5.3 Radioactive Source Term and Lung Absorption Types............................................................. 8

5.3.1 Tritium .......................................................................................................................... 8

5.3.2 Plutonium ................................................................................................................... 10

5.3.3 Uranium ...................................................................................................................... 11

5.3.4 Nickel-63 .................................................................................................................... 12

5.3.5 Carbon-14 .................................................................................................................. 13

5.3.6 Krypton-85.................................................................................................................. 13

5.3.7 Miscellaneous Radionuclides...................................................................................... 13

5.4 Historical Monitoring Practices ............................................................................................... 13

5.4.1 Tritium Contamination Monitoring ............................................................................... 13

5.4.2 Tritium Air Sampling ................................................................................................... 14

5.4.3 Bioassay Monitoring Practices.................................................................................... 15

5.4.3.1 Records Interpretation.................................................................................. 15

5.4.3.2 Tritium Bioassay Monitoring ......................................................................... 15

5.5 In Vitro Bioassay Detection Levels and Analysis Methods...................................................... 17

5.5.1 Tritium Urinalysis Detection Levels............................................................................. 17

5.5.2 Tritium Analysis Methods for in Vitro Urine Samples .................................................. 18

5.6 In Vivo Minimum Detectable Activities and Analysis Methods ................................................ 19

5.7 Interferences and Uncertainties.............................................................................................. 19

5.8 Assessment of Intakes And Doses......................................................................................... 20

5.8.1 Workers Monitored for Soluble Tritium Exposure........................................................ 20

5.8.1.1 Assessment of Soluble Tritium Exposures ................................................... 20

5.8.1.2 Assessment of Potential Unmonitored Exposures to Insoluble Tritium ......... 21

5.8.2 Workers Not Monitored for Soluble Tritium Exposure ................................................. 22

5.9 Incident History ...................................................................................................................... 22

5.10 Attributions and Annotations .................................................................................................. 25

References ........................................................................................................................................ 26

Glossary ............................................................................................................................................ 32
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 4 of 36

LIST OF TABLES

TABLE TITLE PAGE

5-1 Potential tritium radioactive material use areas ...................................................................... 10

5-2 Plutonium radioactive material management areas................................................................ 11

5-3 Default tritium urinalysis MDCs and reporting levels............................................................... 18

5-4 Insoluble tritium intake rates................................................................................................... 22

5-5 Chronology of unusual events and significant activities in relation to internal dosimetry......... 23
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 5 of 36

ACRONYMS AND ABBREVIATIONS 

AEC U.S. Atomic Energy Commission

AWE atomic weapons employer

Bq becquerel

Ci curie

cm centimeter

d days

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

DU depleted uranium

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

hr hour

HT tritium gas

HTO tritium oxide (also known as tritiated water)

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

in. inch

keV kiloelectron-volt, 1,000 electron-volts

L liter

LMSC Lockheed Martin Specialty Components

m meter

MDC minimum detectable concentration

mL milliliter

MMSC Martin Marietta Specialty Components

MPC maximum permissible concentration

MT metal tritide

mrem millirem

NBS National Bureau of Standards

NIOSH National Institute for Occupational Safety and Health

NRC U.S. Nuclear Regulatory Commission

OBT organically bound tritium

ORAU Oak Ridge Associated Universities
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 6 of 36

pCi picocurie

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

RTG radioisotopically-powered thermoelectric generator

SAR safety analysis report

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

Sv sievert

T tritium

T2 tritium gas

T2O tritium oxide

TBD technical basis document

TRS Tritium Recovery System

U.S.C. United States Code

yr year

μCi microcurie

§ section or sections
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 7 of 36

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 historical background information and guidance to assist in the preparation of 

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

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

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

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

of the individual work required for each dose reconstruction.

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

dose reconstructions for covered employees (NIOSH 2010):

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

conventional structures

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

reasons
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 8 of 36

5.2 OVERVIEW

This technical basis document (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. 

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. For convenience, this TBD refers to 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 2011).

The Pinellas 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, highvacuum technology; hermetic seals between glass, ceramic, glass-ceramic, and metal materials; and 

high-voltage generation and measurement technology. The Plant manufactured only neutron 

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

detectors, radioisotopically-powered thermoelectric generators (RTGs), high-vacuum switch tubes, 

specialty capacitors, and specialty 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). 

5.2.1 PURPOSE

The purpose of this 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. 

5.2.2 SCOPE

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

in accordance with OCAS-IG-002, 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. 

5.3 RADIOACTIVE SOURCE TERM AND LUNG ABSORPTION TYPES

5.3.1 TRITIUM

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 
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 9 of 36

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.

Four types of tritium compounds were present at the Pinellas Plant: tritiated water (HTO or T2O), 

tritium gas (HT or T2), organically bound tritium (OBT), and metal tritides (MTs). Based on the 

available records, the predominant tritium exposure hazard to the workers was from HTO and HT. 

However, in some circumstances workers might have also been exposed to OBT compounds and 

MTs. 

The most common OBT compounds at the Pinellas Plant were contaminated pump oils and organic 

solvents (e.g., alcohol, toluene, acetone, etc.) that were used in the tritium areas. Turbo and vacuum 

pumps that were used in systems at the Pinellas Plant periodically required maintenance, which 

included changing the oil. When those pumps were located in tritium areas, the tritium could 

contaminate the pump oil “through small liberations during process operations.” After the oil was 

drained from a pump, a sample of the oil was collected and analyzed for tritium. The nonvolatile 

radioactive liquid wastes not miscible in water, such as the contaminated pump oils, were packaged 

for shipment to an offsite disposal area. The contaminated pump oils were accumulated and stored in 

Area 108 prior to disposal. Based on the available records, organic solvents for cleaning and 

degreasing tritium contaminated components were likely the most common source of tritiumcontaminated organic solvents. The volatile contaminated organic solvents were routinely disposed of 

by evaporation in a hood that was exhausted to the effluent control system and the main stack (Ward 

1973; Barfield and Johnson 1976; DOE 1987; MMSC 1995, p. 35; LMSC 1997).

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 MT uses, only a small 

portion of the worker population had the potential to be exposed to a dispersible form of an MT, and 

those scenarios were typically limited to accidents involving MTs. Titanium tritide was also used in 

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

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

tritide (Burkhart 1990; Eichman 1979; 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. 
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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 Tube testing

131 Final tube test Final tube testing

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 General development

191 Ceramic Product Engineering 

(CPE) hood room

Ceramic product testing facility

191 Radioanalytical laboratory Radiological Laboratory, tritium recovery

194 Engineering environmental testing Engineering environmental testing

200 Test areas Test areas

800 Accelerator and calibration Accelerator and calibration

1000 Waste storage Waste storage

Source: DOE (1995, p. 31); Weaver (1993).

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.

5.3.2 PLUTONIUM

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.

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 undated a). 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 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. 
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Table 5-2. Plutonium radioactive material 

management areas. 

Location Name

400 RTG area

200 Test areas

800 Accelerator

1000 Waste storage

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 source 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 1989). 

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. The available information indicates that Pinellas Plant did not 

receive enough contaminated 238PuO2 heat sources for a worker performing the receipt inspections to 

receive more than a negligible total internal plutonium dose to any organ (i.e. more than 0.001 rem). 

Because of that and because all other sources of plutonium at the Pinellas Plant were in nondispersible forms (e.g. encapsulated sources, electro-plated sources, etc...), it is unlikely that any 

workers at the Pinellas Plant received more than a negligible internal exposure to plutonium, and 

potential plutonium exposures do not need to be assessed for Pinellas Plant workers. As a result, 

plutonium is not discussed any further in this TBD. 

5.3.3 URANIUM

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.
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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 MT 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 

DU (mainly 238U) inside the tritium storage beds presented 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. 

5.3.4 NICKEL-63

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).

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. 
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5.3.5 CARBON-14

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).

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, 

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.

5.3.6 KRYPTON-85

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. 

5.3.7 MISCELLANEOUS RADIONUCLIDES

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. 

5.4 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 at the 

Pinellas Plant to detect potential intakes of radioactive materials. 

5.4.1 TRITIUM CONTAMINATION MONITORING

The monthly Health Physics Reports for the Pinellas Plant indicate that contamination monitoring for 

tritium was performed on a routine basis from the beginning of operations at the Pinellas Plant (GE 

1957–1967, GE 1957–1973, GE 1963, GE 1967). Work areas and personnel were checked for 

contamination on a routine basis. Any significant personnel contamination that could have gone 
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 14 of 36

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

bioassay program. The monthly reports also indicated that whenever contamination levels were 

greater than the contamination control limits, decontamination of those areas was initiated. As early 

as 1959, the contamination control limit for tritium was 2 × 10-5 μCi/in2 (688 dpm/100 cm2

), which was 

for radiologically controlled areas. A circa 1966 smear survey procedure indicated that the tritium 

contamination limit for uncontrolled areas was 220 dpm/100 cm2

, and that the limit for radiologically 

controlled areas had been reduced to 440 dpm/100 cm2 (GE ca. 1966, Burkhart 1989). 

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 1957–1967, GE 1957–1973, GE 1963, GE 

1967). The next highest value was reported in 1959, 1.4 × 106 dpm/100 cm2 (3.3 × 10-2 μCi/in2

) (GE 

1957–1967, GE 1957–1973, GE 1963, GE 1967). The majority of the other reported maximum 

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

Early on, it was known that wet smears (a.k.a. swipes) for tritium contamination were more efficient 

than dry smears. The water geometry for counting was used with the liquid scintillation counter for 

contamination monitoring. Two undated and slightly different tritium contamination smear collection 

and analysis procedures have been found for the Pinellas Plant. The only differences between the 

procedures were the amounts of deionized water being added at the various steps. Both procedures 

utilized a wetted cotton ball for collecting the contamination smear samples. To collect a 

contamination smear sample, 2 or 5 mL of distilled water was pipetted into a paper cup containing a 

cotton ball. The wetted cotton ball was then used to smear the area of interest for contamination 

(usually an area of 100 to 1,000 cm2

), and placed back into the paper cup. The contamination smear 

was then taken to the counting laboratory, where 8 or 10 mL of deionized water was added to the 

cotton ball in the paper cup. The paper cup was then crushed to squeeze out the sample liquid. The 

sample liquid was filtered through a Whatman #1 filter paper before letting it pass into a new paper 

cup. Next, a 0.2-mL aliquot of the filtrate was added to 16.5 mL of liquid scintillation cocktail solution, 

and the sample was counted for 5 minutes (GE 1967–1982). The Whatman #1 filter paper was likely 

used to filter out the large nonrespirable particles from the sample liquid so that the large particles 

would not interfere with the liquid scintillation counting step. The Whatman #1 filter paper has a 

nominal pore size of 11 μm, which would have allowed the respirable tritium particles (i.e., ≤10 μm) to 

remain in the filtrate. By 1977, the sample liquid from tritium contamination smears was no longer 

being filtered (ORAUT 2013a). 

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 1989; LMSC 1995).

5.4.2 TRITIUM AIR SAMPLING

Monitoring for airborne tritium 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 (MPC) 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 
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 15 of 36

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 

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-hour 

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 (NRC) (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. 

5.4.3 BIOASSAY MONITORING PRACTICES

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

performed for. However, plutonium bioassay monitoring is not discussed in this TBD for the reasons 

indicated in Section 5.2.2. Nonroutine bioassay monitoring might have been performed for other 

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

monitoring has been found. 

5.4.3.1 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.

5.4.3.2 Tritium Bioassay Monitoring

Since about 1986, the bioassay program at the Pinellas Plant 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 
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 16 of 36

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 1984b, 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).

The Pinellas Plant 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.

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), International 

Commission on Radiological Protection, 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).

6 36

BW

BW E

dpm min MeV erg mrem 2.22 10 1,440 5.7 10 1.6 10 1,000 μCi day disintegration MeV rem

erg/g 100

rad

V

DCF

M λ QF

− −

∞

× ××

=

                 

     

(5-1)

3

BW

BW

W

L 1 10

mL 42

M

V L

d

  − ×     = = (5-2)

(5-3)

1⁄2

In(2) E

E

λ

T =
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 17 of 36

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)

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). 

( ) ( ) E E 0 (Δ 1

1

λ T λ T D DCF C e C e i i

− − −

∞ ∞ − =   −   (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).

5.5 IN VITRO BIOASSAY DETECTION LEVELS AND ANALYSIS METHODS

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

5.5.1 TRITIUM URINALYSIS DETECTION LEVELS

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

tritiated water, uranium tritide, and certain organically bound tritium compounds). 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, 
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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. 

5.5.2 TRITIUM ANALYSIS METHODS FOR IN VITRO URINE SAMPLES 

Liquid scintillation counting was used to analyze the urine samples for tritium. Urine samples were 

initially analyzed by the Health Physics Laboratory using a Tri-Carb liquid scintillation counter 

(GE 1964; GE 1967–1982, pp. 52–98; Johnson 1977). Based on a sample analysis form from 1964, 

a urine sample prepared for liquid scintillation counting 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 2,5-diphenyloxazole (PPO), and 

0.12 g 1,4-bis[5-phenyloxazol-2-yl]benzene (POPOP) (GE 1964 pp. 89–97). 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, 

1967–1982). 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).

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; Author unknown (1973); GE (1968, 1971, undated a,b,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 the Oak Ridge Associated Universities (ORAU)

Team 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).

By the late-1960s, an analysis procedure indicates that 1 teaspoon of charcoal was being added to 

each urine sample, to decolorize the solution, and the solution was filtered through a Whatman #1 

filter paper before adding the 0.2-mL urine aliquot to the scintillation mix. The individual urine 

samples were counted in the Tri-Carb liquid scintillation counter for 1 minute. If the 1-minute count 

was above the reporting level, the sample was recounted for 5 minutes. The net count rate for the 

sample was determined by subtracting a background count rate before calculating the tritium 

concentration of the urine sample (GE 1967–1982, pp. 52–98). 
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 19 of 36

The background count rates were obtained from uncontaminated urine samples, which were usually 

obtained from new employees or individuals who did not come into contact with radioactivity. Except 

for the count time, the background urine samples appear to have been prepared and analyzed like the 

other urine samples. The count time for the background urine samples was 100 minutes to reduce 

the statistical error. The results from the background urine samples were recorded on a Counter 

Check Log (GE 1967–1982, pp. 72–74). 

By 1973, the Health Physics Laboratory was using a Beckman Model LS-100 liquid scintillation 

spectrometer to analyze the urine samples (Johnson 1977; Ward 1973). On the Beckman counter, 

the preset error control for channel 1 was set to 15%, the preset time was set to 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 0.67-μCi/L reporting level, the result was verified by processing and counting a 

second sample for 5 minutes with a preset error ±5% (GE 1967–1982, pp. 220–221). 

Sometime between 1977 and 1986, the analysis of urine samples transitioned from the Health 

Physics Laboratory to the Environmental Chemistry Laboratory (a.k.a., ECL and Environmental Health 

Section of the Chemistry Laboratory) (Johnson 1977; Molloy 1987). The ECL used two Beckman 

liquid scintillation counting systems, one for high-level radioactivity samples (20 to 1,000,000 cpm) 

and one for low-level radioactivity samples (1 to 30 cpm). Urine samples were evaluated as high-level 

samples. In 1986, the ECL modified and evaluated their two Beckman liquid scintillation counting 

systems to make them more automated and help with the increasing sample analysis loads 

(Molloy 1987). 

By 1977, the ECL was using Handifluor, a premixed scintillation cocktail, for some samples 

(GE 1976–1987). In 1982, Ready-Solv EP, another premixed scintillation cocktail, was evaluated and 

approved for use for in-plant tritium samples; it had a triton/toluene base (Barfield 1982). The use of 

premixed scintillation cocktails eliminated the need for the laboratory to spend time creating those 

mixes. However, as with the earlier cocktails, the premixed cocktails still used flammable organically 

based solvents, which created mixed hazardous waste (i.e., both radioactive and flammable wastes 

for spent tritium samples). In 1986, Opti-Fluor, a water soluble premixed cocktail that uses a higher 

flash point solvent, was evaluated and approved for use as a viable substitute for the more flammable 

cocktails. Tritium contaminated liquid scintillation samples with Opti-Fluor were not considered to be 

mixed hazardous waste when the samples were no longer needed (Klee 1986). 

An undated procedure that appears to be from the late 1960s indicates that daily urine samples from 

multiple workers were composited for screening purposes (GE 1967–1982, pp. 52–98). However, the 

exact use of those screening sample results is uncertain. Based on a review of dosimetry records in 

the Site Research Database (GE 1957 to 1960, 1964, 1966, 1968, 1971, undated b, GE undated c) 

and Pinellas Plant claims, there is no indication of composited sample results being provided in the 

workers’ dosimetry records.

5.6 IN VIVO MINIMUM DETECTABLE ACTIVITIES AND ANALYSIS METHODS

There was no in vivo monitoring at the Pinellas Plant.

5.7 INTERFERENCES AND UNCERTAINTIES

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 

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

in Pinellas Plant procedures (GE undated a). 
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 20 of 36

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 (GE 1986). 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 crosscontamination issue for tritium bioassay samples. 

5.8 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. 

5.8.1 WORKERS MONITORED FOR SOLUBLE TRITIUM EXPOSURE

5.8.1.1 Assessment of Soluble Tritium Exposures

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

HT, HTO, and OBT should be assessed using the worker’s urine sample data. Because there is 

insufficient information to determine what fraction of a worker’s tritium exposure was attributable to 

OBT compounds, both 100% HT and HTO doses and 100% OBT doses should be assessed, and 

then only the most favorable to the claimant dose should be assigned. 

Potential exposures to HT and HTO should be assessed assuming that 100% of the tritium intake was 

attributable to HTO because there is insufficient information to determine what fractions of a worker’s 

tritium exposure were attributable to HT and HTO. Between HT and HTO, HTO is the more 

dosimetrically significant form of tritium. The calculations should be performed using the Excel® 

workbook Tritium Doses from Urine Data Workbook in accordance with the recommendations in 

ORAUT-OTIB-0011, Technical Information Bulletin: Tritium Calculated and Missed Dose Estimates

(ORAUT 2004b). 

Potential exposures to OBT should be assessed using the Integrated Modules for Bioassay Analysis 

(IMBA) computer code, and assuming 100% of the tritium intake is attributable to OBT. The 

calculations should be performed in accordance with the recommendations in ORAUT-OTIB-0060, 

Internal Dose Reconstruction (ORAUT 2014). The following is additional guidance for using the IMBA 

computer code to assess OBT intakes and doses. When the available information indicates that the 

intake mode is inhalation, “Hydrogen-3 (organic)” should be selected for the radionuclide, “Vapor” 

should be selected as the route, then the “SR-2” vapor class option (the only option) should be 

selected in “ICRP Defaults.” When the available information indicates that the intake mode is 

ingestion, “Hydrogen-3 (organic)” should be selected for the radionuclide, “Ingestion” should be 

selected as the route, then the “Ing” absorption option (the only option) should be selected in “ICRP 

Defaults.” When the intake mode is unknown, inhalation should be the assumed intake mode, in 

accordance with ORAUT-OTIB-0060. 
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 21 of 36

5.8.1.2 Assessment of Potential Unmonitored Exposures to Insoluble Tritium

At the Pinellas Plant, insoluble tritium compounds (i.e., certain MTs) were handled only in areas 

where the more dispersible and more soluble forms of tritium (e.g., HT, 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 

insoluble tritium 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.

For the Pinellas Plant, little individual contamination smear data is available. However, the monthly 

Health Physics Reports summarized the status of the radioactive contamination control program and 

reported the highest contamination levels measured for a given month. Therefore, all of the available 

Health Physics Reports (i.e. the various reports for 1957–1973) were reviewed to determine the 

tritium contamination controls at the Pinellas Plant (GE 1957–1967, GE 1957–1973, GE 1963, GE 

1967). These reports indicate that whenever contamination levels were greater than the 

contamination control limits, decontamination was initiated in those areas. The reports also confirmed 

that a routine contamination control program was in place throughout the history of the site and that it 

would have been unlikely for high contamination levels to exist for extended periods. As early as 

1959, the contamination control limit for tritium was 2 × 10-5 μCi/in2 (688 dpm/100 cm2

), which was for 

radiologically controlled areas. A circa 1966 smear survey procedure indicated that the tritium 

contamination limit for uncontrolled areas was 220 dpm/100 cm2

, and that the limit for radiologically 

controlled areas had been reduced to 440 dpm/100 cm2 (GE ca. 1966, Burkhart 1989). An interview 

of a former Pinellas Plant Health Physicist also confirms those radioactive contamination control 

practices. According to him, radiological control personnel would take contamination smears in the 

morning and, if contamination was identified, they would then mop up the area and re-survey. The 

likelihood of routine surface contamination level in the millions of dpm/100 cm2 should be considered 

unusual and short in duration. It was also stated that “metal tritide contamination was cleaned up 

fairly quickly” (ORAUT 2013c). 

The monthly Health Physics Reports also provide information on the maximum tritium surface 

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

4.4 × 106 dpm/100 cm2 (10,000 times the control limit), which occurred in 1970 (ORAUT 2016). The 

next highest value reported was 1.4 × 106 dpm/100 cm2 (3.3 × 10-2 μCi/in2

), which occurred in 1959

(ORAUT 2016). The majority of the rest of the reported annual maximum surface contamination 

levels are at least an order of magnitude lower than these two. However, the following assumptions 

are made to bound the potential unmonitored insoluble tritium exposures: 

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

)

(ORAUT 2016);

• Resuspension factor of 5 × 10-5/m (ORAUT 2013b).

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 with 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 intake rates provided in Table 5-4 should be used to account for 

potential insoluble tritium intakes for the periods when 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.
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 22 of 36

Table 5-4. Insoluble tritium intake rates (ORAUT 2016). 

Hours worked 

per yeara

Annual 

inhalation 

(pCi/yr)

Annual 

ingestionc

(pCi/yr)

Dailyb

inhalation 

(pCi/day)

Daily 

ingestionc

(pCi/day)

2,600 3.09E+07 5.80E+05 8.47E+04 1.59E+03

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 day/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, and then only the most favorable to the 

claimant dose should be assigned.

As indicated above, unmonitored intakes of insoluble tritium will be assessed for all of the periods that 

workers were monitored for soluble tritium exposures. However, in addition to the two bounding 

assumptions mentioned above, there are other reasons that the assessment of those chronic 

unmonitored insoluble tritium intakes will likely result in overestimates of the workers’ exposures to

insoluble tritium. Those reasons include the following:

• the available information on the Pinellas Plant indicates that only a relatively

small portion of the workforce had the potential to be exposed to dispersible

forms of insoluble tritium;

• there is no indication of routine insoluble tritium exposure scenarios (chronic

exposures), and the known exposure scenarios were typically limited to

accidents (acute exposures);

• most of the tritium contamination data that was used for the unmonitored

insoluble tritium intakes were based on results for soluble tritium, which were

also the highest contamination survey results;

• as indicated above, the highest reported surface contamination level was

4.4 × 106 dpm/100 cm2

, which was actually for soluble tritium; whereas, the

highest known insoluble tritium contamination level was only 1.89 × 105 dpm/100

cm2 (i.e. only about 4% of the surface contamination level used to estimate the

unmonitored insoluble tritium intakes) (GE 1957–1973).

Even though the unmonitored intakes above likely result overestimates of the workers’ exposures to 

insoluble tritium, those unmonitored intakes are still considered to be reasonable estimates.

5.8.2 WORKERS NOT MONITORED FOR SOLUBLE TRITIUM EXPOSURE

Workers at the Pinellas Plant were likely only exposed to onsite levels of environmental radioactivity 

during the unmonitored periods of their employment. 

5.9 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. 03 Effective Date: 07/18/2016 Page 23 of 36

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 leaking uranium storage 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). 

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 Not 

applicable

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 Not 

applicable

02/21/59 Area contamination – operator broke glass system Not reported

03/12/59 Operator contaminated during system cleaning by another worker (Room 14) Not reported

06/04/59 Personnel error working on Stack Effluent Control System (SECS) test system –

Room 21

753

06/05/59 Area contamination-diffusion pump exploded in Hood 14 Not reported

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 Not reported

05/14/60 Broken flask caused area contamination – Room 10 Not reported

06/21/60 Ion gauge exploded – Room 16 Not reported

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

07/13/60 Manifold shattered, exposing worker – Room 23 Not reported

08/12/60 Contamination spread TiH2 in Area 108 from broken flask Not reported

09/10/60 Requirement for full anticontamination clothing in Area 108 reduced to laboratory

coats for normal production operations

Not reported

11/14/60 Began using NBS Handbook 69 (NBS 1959) for MPCs Not 

applicable

04/61 Area contamination from system breakage Not reported

10/62 A modified personal monitor was installed in Area 108 Not 

applicable

12/62 Breathing air supply line connected to Area 108 exhaust duct Not 

applicable

09/63 Employees found falsely identifying urine samples Not 

applicable

03/30/65 Broken flask – Room 9 Not reported

03/20/65 Flask explosion – Room 12 Not reported

05/66 SECS cold water removal problems 252

01/27/67 Glovebox vacuum pump oil degassed 32

10/12/67 Personnel contamination – O-ring mishandled – Room 18 Not reported

06/18/68 Acid cleaning explosion – Area 181 Not reported

02/69 Leaking flange at absorption pump in Area 108 8
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 24 of 36

Date Description

Curies of 

tritium 

released

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 Not reported

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

moved to burning pad west of 400.

Not reported

01/70 Area contamination and personnel exposure from flaking tube part in gas lab Not reported

02/70 Area contamination from pressurized sorb pump (air expansion) – Room 2 Not reported

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

Area 182D

Not reported

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 tritium-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

05/24/73 Areas 156, 157, and 158b contaminated with tritium due to an MT particle spill in 

Area 158b. The MT particles were from a zirconium/aluminum getter.

Not reported

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-in. valve 0

04/13/76 Oven fan blade broke tubes 0

02/77 Packaging fixtures in Area 182D glovebox 28

09/11/79 Work in Room 18 hood – internal dose 5.6

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

08/80 Contaminated electron microscope 0

04/81 Three waste drums found outgassing during truck loading were removed to 

Area 108; a procedure was put in place to check all drums for outgassing before

transfer to the storage building. 

Not reported

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

04/20/82 TRS valve left 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 Not 

applicable

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 Laboratory 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

On [Redacted], a release of 5.6 Ci of HTO from the main exhaust stack occurred at approximately 

[Redacted]. The release resulted from nonroutine modifications being made to the [Redacted] system 

in [Redacted] of [Redacted]. Analyses of urine samples from the technician performing the work 
Document No. ORAUT-TKBS-0029-5 Revision No. 03 Effective Date: 07/18/2016 Page 25 of 36

indicated that the worker received a body burden of 62.2 μCi/L, which the Pinellas Plant equated to a 

dose of 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 problem was identified due to the stack 

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

[Redacted] of [Redacted] (Maurer 1979).

In 1995, several leak standards (cylinders containing hydrogen, Freon, and argon), 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 the Pinellas Plant conducted a contamination 

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

Each of the above incidents indicates that the Pinellas Plant 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 plant operations in 1957.

5.10 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. 03 Effective Date: 07/18/2016 Page 26 of 36

REFERENCES

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Company, Pinellas Plant, St. Petersburg, Florida, May 19. [SRDB Ref ID: 13125]
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Ward, W. E., 1971, “Sensitivity Levels of Radiation Measurements,” memorandum to E. P. Forest, 

General Electric Company, Pinellas Plant, St. Petersburg, Florida, February 5. [SRDB Ref ID: 

12833, p. 10-11]

Ward, W. E., 1973, Tritium Control Technology, GEPP-104, General Electric Company, Pinellas 

Plant, St. Petersburg, Florida, July 16. [SRDB Ref ID: 13128, p. 15-40]

Weaver, A. S., 1989, “1988 Plutonium Bioassay Results,” memorandum to file, General Electric 

Company, Pinellas Plant, St. Petersburg, Florida, January 27. [SRDB Ref ID: 12743, p. 6–7]

Weaver, A. S., 1990, External Dosimetry, HP03, Rev. 4.0, General Electric Company, Pinellas Plant, 

St. Petersburg, Florida, June 1. [SRDB Ref ID: 12989, p. 2-6]

Weaver, A. S., 1992, Summary: Natural Uranium Glass Concerns, General Electric Company, 

Pinellas Plant, St. Petersburg, Florida, January 31. [SRDB Ref ID: 13204]

Weaver, A. S., 1993, Pinellas Plant Radiation Safety Overview and Visitor Orientation, Martin Marietta 

Specialty Components, Pinellas Plant, Largo, Florida, February 22. [SRDB Ref ID: 12814]

Weaver, A. S., 1994, Health Physics Consensus Statement #5, Pinellas Plant Internal Dosimetry 

Program, Lockheed Martin Specialty Components, Pinellas Plant, St. Petersburg, Florida, 

November 22. [SRDB Ref ID: 12165, p. 13-14]

Weaver, A. S., ca. 1995, Internal Dosimetry Questions, Pinellas Plant, Martin Marietta Specialty 

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GLOSSARY

absorption type

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). Also called solubility type.

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 radionuclide material in the body (in vivo 

measurement) or in biological material excreted or removed from the body (in vitro 

measurement) and analyzed for purposes of estimating the quantity of radioactive material in 

the body. Also called radiobioassay. 

bioassay procedure

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

Amount of radioactive material in an individual's body at a particular point in time.

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.

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. 

dose equivalent

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

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

from the absorbed dose. See dose. 
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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.

intake

Radioactive material taken into the body by inhalation, absorption through the skin, injection, 

ingestion, or through wounds.

internal dose or exposure

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

sources).

internal dose assessment

Estimation of an intake of radioactive material and the consequent radiation dose based on 

bioassay or other measurements in the work environment.

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

Measurements 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.

kiloelectron-volt (keV)

Unit of particle energy equal to 1,000 (1 × 103

) electron-volts.

limit of detection

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.
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lung absorption type

See absorption type.

metal tritide (MT)

Metal 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 activity or amount

Smallest amount (activity or mass) of an analyte in a sample that can be detected with a 

probability β of nondetection (Type II error) while accepting a probability α of erroneously 

deciding that a positive (nonzero) quantity of analyte is present in an appropriate blank sample 

(Type I error).

minimum detectable concentration (MDC)

Minimum detectable activity (or amount) in units of concentration. See minimum detectable 

activity. 

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

Internal and external ionizing radiation dose from exposure during employment. Occupational 

dose does not include that from background radiation or medical diagnostics, research, or 

treatment, but does include dose from occupationally required radiographic examinations that 

were part of medical screening.

organically bound tritium (OBT)

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. 
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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.

sievert

International System unit for dose equivalent, which indicates the biological damage caused 

by radiation. The unit is the radiation value in gray (equal to 1 joule per kilogram) multiplied by 

a weighting factor for the type of radiation and a weighting factor for the tissue; 1 sievert 

equals 100 rem.

soluble tritium

All forms of tritium, except for those that have type M or S lung absorption properties.

special monitoring

Monitoring in addition to the routine monitoring program carried out for special purposes such 

as estimating the amount of radionuclide deposited in a person after a known or suspected 

accidental intake or after a known or suspected environmental release. See routine 

monitoring. 

spot sample

In relation to bioassay, a single void of urine.
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tritium

Radioactive isotope of hydrogen that contains one proton and two neutrons in its nucleus. It 

decays by beta emission and has a radioactive half-life of about 12.5 years.

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.