ORAU TEAM

Dose Reconstruction

Project for NIOSH

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

Page 1 of 31

Document Title:

Pinellas Plant – Occupational Internal 

Dose 

Document Number: ORAUT-TKBS-0029-5 

Revision: 00

Effective Date: 10/07/2005

Type of Document: TBD

Supersedes: None

Subject Expert: Paul J. Demopoulos

Document Owner

Approval: Signature on File Approval Date: 09/28/2005

Vernon S. McDougall, TBD Team Leader

Approval: Signature on File Approval Date: 09/28/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Concurrence Date: 09/28/2005

Richard E. Toohey, Project Director

Approval: Signature on File Approval Date: 10/07/2005

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

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

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 2 of 31

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.
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 3 of 31

TABLE OF CONTENTS

SECTION TITLE

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

PAGE

5

5.1 Introduction.................................................................................................................................6

5.2 In Vitro Minimum Detectable Concentrations And Counting Methods........................................6

 5.2.1 Plutonium Minimum Detectable Concentration Determination 

 and Counting Methods................................................................................................. 6

 5.2.2 Tritium Minimum Detectable Concentrations ................................................................8

 5.2.3 In Vitro Methods for Individual Radionuclides ..............................................................9

 5.2.3.1 In Vitro Bioassay for Tritium........................................................................... 9

 5.2.3.2 In Vitro Bioassay for Plutonium.................................................................... 11

 5.2.4 Frequency of Urine Bioassays....................................................................................11

 5.2.4.1 Frequency for tritium oxide (HTO) Urine Bioassays ...................................... 11

 5.2.4.2 Frequency for Plutonium Urine Bioassays .................................................... 12

 5.2.5 Urinalysis Minimum Detectable Concentration Summary..................................12

5.3 Source Term and Isotopic Determination – Swipes and Air Sampling......................................12

 5.3.1 Contamination Monitoring...........................................................................................12

 5.3.1.1 Tritium Oxide Contamination Monitoring ..................................................... 12

 5.3.1.2 Plutonium Contamination Monitoring .......................................................... 13

 5.3.1.3 Nickel Contamination Monitoring ................................................................ 14

 5.3.1.4 Carbon-14 Contamination Monitoring ......................................................... 14

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

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

 5.3.2.2 Plutonium Air Sampling................................................................................ 14

5.4 Urinalysis Volume.....................................................................................................................15

5.5 Absorption Types for Radionuclides Included in the Bioassay Program...................................15

5.6 Radionuclides Not Included in the Bioassay Program..............................................................16

5.7 In Vivo Minimum Detectable Activities, Counting Methods, and Reporting Practices..............16

5.8 Interference and Uncertainty.....................................................................................................16

 5.8.1 Uncertainty and the Contamination of Samples............................................... 16

5.9 Assessment of Intake for Monitored and Unmonitored Workers...............................................17

 5.9.1 Monitored Workers with Measurable Intakes (HTO and

 Plutonium).. .....................................................................................................17

 5.9.2 Metal Tritide Exposures ..................................................................................17

5.10 Tritium Dose Distribution information........................................................................................18

 5.10.1 Relative Biological Effectiveness for Tritium....................................................20

5.11 In Vitro Urinalysis Counting Form Code Information.................................................................21
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5.12 Tritium and Plutonium Potential Radioactive Material Use Areas............................................21

5.13 Incident History.........................................................................................................................22

References....................... ................................................................................................................. 25

Glossary................ ............................................................................................................................29

LIST OF TABLES

TABLE TITLE PAGE

5-1 In vitro measured concentrations or MDC averages for plutonium 

urinalysis by year .....................................................................................................................9

5-2 In vitro measurement frequencies, measurement types and MDCs for 

tritium urinalysis .....................................................................................................................10

5-3 In vitro measurement frequencies, measurement types, and MDCs for 

urinalysis................................................................................................................................12

5-4 Job Titles NOT Likely to Receive MT Exposures.................................................................... 17

5-5 Summary of tritium urinalysis from 1959 to 1962.................................................................... 18

5-6 Summary of tritium urinalysis from 1986 to 1991.................................................................... 19

5-7 Summary of average and maximum tritium doses derived from tritium 

urinalysis by year ................................................................................................................... 19

5-8 Potential tritium radioactive material areas............................................................................. 21

5-9 Plutonium radioactive material management areas ................................................................ 22

5-10 Chronology of unusual events and significant activities in relation to 

internal dosimetry................................................................................................................... 23
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ACRONYMS AND ABBREVIATIONS

AMAD activity median aerodynamic diameter

AEC U.S. Atomic Energy Commission

CF calibration factor

cm centimeter

cpm counts per minute

DIL derived investigation level

DOE U.S. Department of Energy

dpm disintegrations per minute

DU depleted uranium

EDE effective dose equivalent

g gram

HEPA high-efficiency particulate air

hr hour

ICRP International Commission on Radiological Protection

keV kilovolt-electron, 1,000 electron volts

HTO Tritiated water vapor 

L liter

LSC liquid scintillation counting

MDC minimum detectable concentration

MDL minimum detectable level

MeV megavolt-electron, 1 million electron volts

min minute

ml milliliter

mR milliroentgen

MT Metal tritide

mrad millirad

mrem millirem

NIST National Institute of Standards and Technology

RBE relative biological effectiveness

ROI region of interest

T Tritium

U.S.C. United States Code

μCi microcurie

μm micrometer
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 6 of 31

5.1

Technical Basis Documents and Site Profile Documents are general working documents that provide 

guidance concerning the preparation of dose reconstructions at particular sites or categories of sites. 

They will be revised in the event additional relevant information is obtained about the affected site(s). 

These documents may be used to assist the National Institute for Occupational Safety and Health in 

the completion of the individual work required for each dose reconstruction.

INTRODUCTION

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

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

employer facility” or a “Department of Energy (DOE) facility” as defined in the Energy Employees 

Occupational Illness Compensation Program Act of 2000 [42 U.S.C. Sections 7384l(5) and (12)]. 

The Pinellas Plant internal dosimetry program started with site operations in 1957. Tritium “(T)” and

238Pu and 239Pu were the isotopes of primary concern and the only radionuclides monitored routinely. 

The primary method for monitoring of employees for radionuclide intake was urine bioassay. Pinellas 

used 14C and 63Ni, but in much smaller quantities; therefore, Pinellas did not perform bioassays for 

these radionuclides.

5.2 IN VITRO MINIMUM DETECTABLE CONCENTRATIONS AND COUNTING METHODS

5.2.1

The minimum detectable concentration (MDC) calculation was based on the formulation in Equation 

5-1 for the Pinellas plutonium urinalysis program conducted from 1975 through 1990. The Pinellas 

facility used MDCs as the decision levels. Pinellas labeled the MDC as MDL (minimum detectable 

level). Because the following formulation was used for the MDL calculation, if the background counts 

were zero, then the MDL result was zero [this seldom occurred (PDN 240001295)].

Plutonium Minimum Detectable Concentration Determination and Counting Methods

MDC = 4.65 × (CBackground)

1/2 × CF

TREVA (5-1)

where

MDC = minimum decision level (μCi/L)

CBackground = background counts in the region of interest of either 238Pu or 239Pu (from 1988 and 

forward) or peak centroid (from 1975 to 1987)

T = count time 

R = recovery fraction

E = average detector efficiency

V = sample volume (L), and

A = the alpha abundance for the radionuclide in question

CF = (Calibration factor) 242Pu concentration × 242Pu volume (from at least 1988 to 

present)

The MDC was the level at which activity was considered present in a sample with a 95% confidence 

level. 
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 7 of 31

Plutonium MDCs for either 238Pu or 239Pu was calculated with the following formula:

APu = {Net ROICTS (238Pu or 239Pu) × CF} (5-2)

{Net ROICTS (242Pu) × sample size} 

From 1975 through 1982 the determination of a positive count for plutonium was not specifically 

discussed in the Pinellas Plant documents. In 1983 two bioassay samples equaled the MDLs: One 

for 238Pu, where the activity equaled the MDL of 9.66 × 10-12 μCi/ml, and one for 238Pu, where the 

activity equaled the MDL of 1.82 × 10-11 μCi/ml. One other bioassay found a concentration below the 

MDL for 238Pu and was recounted and found with no positive activity (PDN 240000887).

The two samples from 1983 mentioned above were recounted five times each. In the first case two of 

the recounts for 238Pu activity were one-half of the MDL (which should have been the decision level), 

one was equal to the MDL, and two found zero activity. In the second case four of the results found 

zero activity, and one was equal to the MDL. It is apparent after these recounts that no positive 

activity was expected (PDN 240000887). (No dose calculations were ever performed at Pinellas for 

plutonium except under a fictitious worst-case incident scenario.)

In 1988, more specific requirements for the determination of a positive activity sample were 

developed. Five separate criteria had to be met for the determination of a positive sample. The five 

criteria were (Weaver 1989): 

1. The ratio of 238Pu to 239Pu in a bioassay sample must be about 5:1 (±20%) or 238Pu is detected 

while 239Pu is not detectable.

2. The result must be greater than the MDL at the 95% confidence level.

3. The precision of the result must be known at 3 sigma (Pinellas used 2 sigma as shown in the 

bioassay documents) so that the negative range does not include zero.

4. The 238Pu value is above the Derived Investigation Level (DIL) of 3.23 × 10-12 μCi/ml or the 

239Pu value is above the DIL of 2.86 × 10-12 μCi/ml. Calculated DILs are for inhalation 

pathway, class Y, particle size 1-μm activity median aerodynamic diameter (AMAD), and a 

yearly sample.

5. The recovery of the tracer must be greater than 50%. 

If the first four criteria are met and the fifth is not, a new sample will be processed before 

further actions are taken.

The bioassay samples were typically counted for about 72 hr, but the actual count times may have 

varied from 24 to 72 hours. When an unexpected result occurred, Pinellas health physicists 

requested that a sample be counted at least 6 days to improve statistics.

If 239Pu was detected above the limit in question (MDL, DIL, etc.) in the absence of 238Pu, the source 

of exposure was investigated but it was not considered a true occupational dose unless otherwise 

proven.

If the five criteria were met for a bioassay sample, a complete investigation was conducted, which 

would include additional urine samples, estimation of dose equivalent (the error would be large using 

only one data point), and a complete survey of Building 400. 
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 8 of 31

No confirmatory evidence of a positive plutonium bioassay or a dose calculation and investigation was 

found in the Pinellas Plant documentation. 

Pinellas health physicists analyzed the 16 plutonium bioassay samples for 1988 using the five criteria 

and the stipulations and determined that there were no positive samples. A summary of the analysis 

was given in the 1988 report. Ten of the 16 samples were considered not positive because the 

238Pu/239Pu ratio was less than about 5:1, five samples had values less than the MDLs for both 238Pu 

and 239Pu, and one sample had both the “3 sigma not precise” and a 238Pu/239Pu ratio of less than 5:1 

(Weaver 1989).

If the same criteria were applied to 1989 and 1990 bioassay data, the following would result:

• The actual 16 bioassay samples for 1989 were analyzed using the five criteria and the 

stipulations and it was determined that there were no positive samples. Six of the 16 samples 

were considered not positive because of zero activity, seven samples had values of zero or 

less than the MDLs for 238Pu and 239Pu, two samples had a 238Pu/239Pu ratio less than 5:1, and 

one had a zero MDL reported for one radionuclide and zero activity for the second 

radionuclide (PDN 240001295).

• The actual 12 bioassay samples for 1990 were analyzed using the five criteria and the 

stipulations and it was determined that there were no positive samples. Four of twelve 

samples were considered not positive because of zero activity, six of twelve samples had 

values of zero or less than the MDLs for both or either 238Pu and 239Pu, one sample had a 

238Pu/239Pu ratio less than 5:1, and one samples had a zero MDL reported for one radionuclide 

and zero activity for the second radionuclide (PDN 240001295).

Table 5-1 lists the average and maximum MDCs and plutonium concentrations in μCi/ml by year. 

Data for 1984 was not found.

It is reasonable to assume that there were no plutonium intakes at the Pinellas plant because:

1. There were no documented plutonium contaminations.

2. There were no documented cases of plutonium air contamination.

3. There were no plutonium-related incidents documented. 

4. All potentially maximally exposed personnel (RTG workers) were routinely monitored for 

plutonium intakes during the time period of RTG manufacture.

5.2.2

Table 5-2 shows that the tritium MDCs varied over the years at Pinellas Plant. The recording levels 

and dose calculations also changed over time. The MDCs were based on the environmental water 

sample geometry or the counting methods of earlier years until about the mid-1970s (GE no date). 

Individual worker results should be used because the stated recording levels as listed in Table 5-2 

varied in the actual records. 

Tritium Minimum Detectable Concentrations 
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Table 5-1. In vitro measured concentrations or MDC averages for plutonium urinalysis by year. 

Period Radionuclide

Average 

MDC 

μCi/ml

Maximum 

MDC 

μCi/ml

Average activity 

concentration 

μCi/ml

Maximum activity 

concentration μCi/ml Reference

1975 Pu-238 2.79E-11 7.88E-11 PDN 240000887

Pu-239 1.66E-11 3.00E-11 PDN 240000887

1976 Pu-238 3.85E-11 1.79E-10 PDN 240000887

Pu-239 1.84E-11 6.86E-11 PDN 240000887

1977 Pu-238 6.79E-12 1.34E-11 3.38E-12 3.38E-12 PDN 240000887

Pu-239 6.03E-12 1.09E-11 1.29E-11 4.93E-11 PDN 240000887

1978 Pu-238 2.77E-11 8.02E-11 0.00E+00 0.00E+00 PDN 240000887

Pu-239 2.77E-11 8.02E-11 1.51E-11 4.01E-11 PDN 240000887

1979 Pu-238 3.64E-12 8.72E-12 0.00E+00 00E+00 PDN 240000887

Pu-239 3.09E-12 8.89E-12 3.79E-12 5.51E-12 PDN 240000887

1980 Pu-238 6.23E-11 3.17E-10 3.84E-12 3.85E-12 PDN 240000887

Pu-239 3.41E-11 1.90E-10 3.34E-12 3.34E-12 PDN 240000887

1981 Pu-238 5.65E-13 2.31E-12 0.00E+00 0.00E+00 PDN 240000887

Pu-239 1.70E-13 4.72E-13 0.00E+00 0.00E+00 PDN 240000887

1982 Pu-238 9.86E-12 1.98E-11 9.88E-12 9.88E-12 PDN 240000887

Pu-239 9.81E-12 1.98E-11 0.00E+00 0.00E+00 PDN 240000887

1983 Pu-238 1.74E-11 3.83E-11 8.13E-12 1.82E-11 PDN 240000887

Pu-239 1.70E-11 3.83E-11 6.88E-12 9.09E-12 PDN 240000887

1985 Pu-238 4.76E-12 8.31E-12 4.76E-12 8.31E-12 PDN 240000887

Pu-239 2.42E-11 2.21E-10 0.00E+00 0.00E+00 PDN 240000887

1986 Pu-238 1.27E-11 3.12E-11 4.13E-12 4.13E-12 PDN 240000887

Pu-239 1.51E-11 3.23E-11 0.00E+00 0.00E+00 PDN 240000887

1987 Pu-238 7.95E-12 1.88E-11 6.70E-12 1.71E-11 PDN 240001295

Pu-239 8.00E-12 1.88E-11 6.25E-12 9.11E-12 PDN 240001295

1988 Pu-238 1.25E-11 3.71E-11 1.06E-11 8.85E-11 PDN 240001295

Pu-239 7.69E-12 3.17E-11 6.41E-12 3.47E-11 PDN 240001295

1989 Pu-238 3.87E-11 1.14E-10 1.35E-11 4.03E-11 PDN 240001295

Pu-239 3.40E-11 1.27E-10 9.94E-12 3.25E-11 PDN 240001295

1990 Pu-238 2.37E-11 3.12E-11 6.88E-12 1.69E-11 PDN 240001295

Pu-239 1.91E-11 3.94E-11 2.25E-12 8.28E-12 PDN 240001295

5.2.3

5.2.3.1 In Vitro Bioassay for Tritium

In Vitro Methods for Individual Radionuclides

Beginning at least as early as 1966, urine samples for liquid scintillation counting (LSC) were 

submitted on a daily basis at the end of a worker’s shifts or at the beginning of the next shift for each 

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

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

sample that was then sampled and counted. A sample consisted of 0.2 ml of the composite sample 

combined with 16.5 ml of fluour (1 L toluene, 0.2 L ethanol, 3.5 g PPO, 0.12 g) (GE 1964). The MDC 

for the samples should have been the MDC for water samples or 0.09 μCi/L, but a value of 0.67 μCi/L 

was used as the recording level and listed in some documentation (Author unknown 1973). A chronic 

infinite dose of 5 mrem whole-body dose was calculated at this 0.67 μCi/L excretion level 

(GE no date).
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Table 5-2. In vitro measurement frequencies, measurement types and MDCs for tritium urinalysis.

a,b

Period Frequencyd MDCμCi/L

Recording level

μCi/L Dose record level Reference

1957 Weekly 0.30 <1 μCi/L <5 mrem

1 μCi/L = 6 mrem

Claims and PDN 240001043

1958 Weekly 0.30 <1 μCi/L <5 mrem

1 μCi/L = 10 mrem

Claims #13872

1959 Weekly 0.30 0.01 μCi/L <1 mrem

1 μCi/L = 10 mrem

Claims #14194

1960 Weekly 0.30 0.01 μCi/L <20 mrem or < Claims #14194

1961 Weekly 0.30 0.01 μCi/L <20 mrem or < Claims #14194

1962 Weekly 0.13 0.67 or < μCi/L <20 mrem or < Claims #14194

1963 Weekly 0.13 0.67 or < μCi/L <20 mrem or < Claims #14194

1964 Weekly 0.13 0.67 or < μCi/L <5 mrem ? Claims #14194

1965-1972 Weekly 0.13 0.67 or < μCi/L <5 mrem – 1966 Claims and GE 1971

1973-2/1974 Daily to monthly? 0.13 0.67 μCi/La = 5 mrem 

infinite dose 1,250 

mrem/yr

<5 mrem ? GE 1973

3/1974-1986 Daily to monthly? 0.10 +/- 0.01 > or = 0.1 μCi/L (0.42 

mrem)

1978 

1 μCi/L = 5 mrem 

Claims #14194; GE 1975; 

GE 1984a

1987-1989 Daily to monthly 0.01 0.10 μCi/L (5 

mrem/μCi/L infinite 

dose – 12-day eff. 

T1/2c)

1987 

4.21 mrem/μCi/L 

infinite dose – 

10-day eff. T1/2c

PDN 240000969

1990-1997 Daily to monthly 0.006 0.10 μCi/L 4.21 mrem/μCi/L 

infinite dose – 

10-day eff. T1/2c

PDN 030000352

PDN240001222

a. Sources: Radioanalyses Sensitivities (GE 1973); Holliday (1983); Burkhart and Richardson (1986).

b. Urinalysis method was LSC for tritium oxide (HTO) throughout Pinellas Plant site history.

c. eff. T1/2 = effective half-life.

d. See section 5.2.4.1 for more detailed frequency information.

A procedure dating from the 1960s was first to add charcoal to the solution to decolorize the solution, 

filter the composite through Whatman #1 filter paper, and pipette 0.2 ml of sample (with the addition of 

the 16.5 ml of cocktail solution). The composite sample was then placed in the freezer for cooling and 

counted for 10 minutes on a Beckman multichannel LSC using the first channel set at 15% preset 

error. If the composite sample exceeded an integrated infinite dose of 200 mrem for 40 samples or a 

5-mrem average infinite dose, then the composite samples were broken into four samples. The 

positive group of samples was than identified and counted individually. Individual samples at this time 

were counted for 5 minutes (GE no date, pp. 35–37). This compositing of urine samples was likely 

used for a screening purpose to reduce the amount of urine samples analyzed and to address 

possible high intakes more quickly. The composite samples should not affect the actual dose 

calculations that the dose reconstructor performs.

Beginning at least as early as 1986, urine samples were submitted at least 1.5 to 2 hr after suspected 

exposure. Participants were instructed to leave at least a 25 ml sample for analysis. The bioassay 

samples were loaded into the LSC, usually overnight, to count for 30 minutes and or to a ±10% error, 

whichever occurred first. One ml of urine was added to 9 ml of LSC cocktail. The MDL of 0.006 μCi/L 

was calculated based on the worst-case background and efficiency values (Burkhart 1995a).

Typical efficiency values for the LSC for tritium in 1957 were about 5% to 8%. The efficiency 

increased to a maximum of about 40% by the 1990s. 
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5.2.3.2 In Vitro Bioassay for Plutonium

The sample was prepared with a number of steps. The first step was a 1-day dissolution process. 

The second step was a solution purification with the use of an anion exchange resin column. The 

third step was a 1-day electroplating. The final step was alpha spectroscopy (Christy 1988).

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

from a one-channel analysis to an energy region of interest (ROI) technique. Pinellas used National 

Institute of Standards and Technology (NIST)-traceable sources of 238Pu and 239Pu to set up the ROIs. 

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

At an unknown point, the Pinellas chemistry counting laboratory used a Series 100 Canberra alpha 

spectroscopy system. This alpha spectroscopy system likely used a silicon barrier detector system 

[as shown in a sample spectrum comparison to a NIST 238Pu and 239Pu source spectrums (GE 1989b, 

pp. 1–3)]. The energy ROI for 238Pu was 5,409.7 to 5,535.5 keV and that for 239Pu was 5,063.4 to 

5,186.6 keV. The 1988 and 1989 bioassay result folders (PDN 240001295) show that the ROIs 

varied. The 200-keV interval between the ROIs for 238Pu and 239Pu indicates that very good resolution 

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

24-hr counting time was used. A bioassay recount was made for 6 days or 144 hr to improve 

counting statistics. 

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

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

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

compared favorably (Burkhart 1990). 

5.2.4

5.2.4.1 Frequency for tritium oxide (HTO) Urine Bioassays

Frequency of Urine Bioassays

The bioassay program at Pinellas since about 1986 was based upon ANSI standard N13.14-1983 

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

included:

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

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

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

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

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

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

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

stated in the Pinellas internal dosimetry technical basis document (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)
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 12 of 31

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)

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

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

Plant.

5.2.4.2 Frequency for Plutonium Urine Bioassays

Plutonium bioassays were performed on an annual basis for all radioisotopic thermal generator (RTG) 

workers from 1975 to 1990. Termination bioassay results that would have been submitted early in 

1991 were not found. 

5.2.5

There have not been a variety of methods used historically to analyze for tritium or plutonium at 

Pinellas. LSC was always used for tritium analysis, and alpha spectroscopy after separation 

chemistry was always used for plutonium analysis. Table 5-3 summarizes these methods and their 

associated detection capabilities. Radionuclides that were not included in the bioassay program at 

Pinellas but were used at unknown quantities are included in Table 5-3. 

Urinalysis Minimum Detectable Concentration Summary

Table 5-3. In vitro measurement frequencies, measurement types, and MDCs for urinalysis.

a

Period Frequency Radionuclide MDC Record level Urinalysis method Restriction Volume

1957-

1997

Daily to monthly, 

based on work 

assignments, 4 hr after 

work permit, daily for 

routine workers, 

weekly or monthly for 

nonroutine workers. 

HTO 0.30 -

0.006 

μCi/L

0.01 - < 1μCi/L LSC – 2 hr wait for a 25 ml 

sample. 2 ml maximum 

combined with 8 to about 

15 ml LSC cocktail.

10 μCi/L (1986-

1995?)

Spot – 25-ml 

sample 

composited, 

then individual

1975-

1990

Annual, for all RTG 

workers 

Pu-238

Pu-239

See 

records

See 

Table 

5-1 

NA -See text. Alpha Spectrometry 

Procedure changed at 

least once sometime in 

1987, see records.

Never occurred at 

Pinellas; only if

sample considered 

positive.

24-hr sample 

with use of 800 

ml maximum or 

whatever was 

available.

a. Sources: Radioanalyses Sensitivities (GE 1973); Holliday (1983); Burkhart and Richardson (1986).

5.3 SOURCE TERM AND ISOTOPIC DETERMINATION – SWIPES AND AIR SAMPLING 

5.3.1

5.3.1.1 Tritium Oxide Contamination Monitoring

Contamination Monitoring

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

operations at the Pinellas Plant. Work areas and personnel were checked for contamination on a

routine basis. Any significant personnel contamination that could have gone undetected from 

contamination surveys would most likely have been identified through the tritium bioassay program. It 

was known early that wet swipes for contamination were more efficient than dry swipes. The water 
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 13 of 31

geometry for counting was used with the LSC for contamination monitoring. This same water 

geometry was used for HTO until about 1971 for the urine bioassays as well (GE no date).

Contamination monitoring during 1962 indicated levels of about 1.3 × 105 dpm/100 cm2 (or up to 

1,000 cm2

) in the work areas with a maximum of 7.8 × 105 dpm/100 cm2 (or up to 1,000 cm2). 

Personal skin contamination averaged about 5 × 105 dpm/100 cm2 (or up to 1,000 cm2

) with a 

maximum of 2.75 × 107 dpm/100 cm2 (or up to 1,000 cm2

) (Jech 1963a). Contamination monitoring 

from 1960 indicated levels of up to 0.352 μCi in the hood systems with 23 μCi in the hoods and up to 

0.043 μCi on the floors. The maximum internal tritium exposure in 1 month was 34 mrem for 1960 

[Monthly Statistics and Summary 1960 (PDN 240001051)].

The procedure for collecting the smears involved pipetting 5 ml of distilled water containing the cotton 

ball swipe, taking the wet cotton ball to the area, and swiping the area of concern (usually of 100 to 

1,000 cm2

). Upon return to the counting lab, 10 ml of water was added to the swipe container, and 

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

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

0.002 μCi/area swiped (GE no date). As early as 1966 the uncontrolled area limit for tritium was 220 

dpm/100 cm2 and the controlled area limit was 440 dpm/100 cm2 (GE c. 1966).

In the 1990s 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 (GE 1989b; GE 1995).

5.3.1.2 Plutonium Contamination Monitoring

From the beginning of plutonium source use in the RTG in 1975, the Pinellas Plant routinely 

monitored the work areas for possible plutonium contamination. This occurred despite the fact that 

the sources were triply encapsulated and were held to more stringent standards than other Federal 

regulatory agencies required for sealed sources of radiation. 

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

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

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

Upon receipt of plutonium sources directly to Building 400, there were leak checks in the receiving 

area in the containment hood, and the sources were repackaged into 2R storage and transport 

containers and placed into the storage vault. A 6-dpm/swipe release limit was the control level for the 

source contamination release limit at the 95% confidence level [an Eberline SAC-4, ESP-1/Ludlum 

43-10 or Baird accu-count was used (Burkhart 1989)]. A maximum of 1 of 70 plutonium sources 

received during the course of one month had a contamination level greater than the detection limit of 

6 dpm. All were decontaminated before RTG processing. Plutonium sources greater than 200 dpm 

contamination levels were sent back to the vendor (Huffman 1979). There was no documentation 

found indicating that any units had to be sent back.

When received in the glovebox area for processing while in a fume hood, the plutonium sources were 

checked with a PAC 4S portable alpha monitor, and the local area was monitored with an Eberline 

alpha-1, -2, -3, or -5 air monitor. This occurred after the user donned a half-face respirator (Type H 

cartridges), laboratory coat, and gloves. The PAC 4S had a 50 dpm/probe lower limit of detection 

(Burkhart 1989). 
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 14 of 31

Shipments of 238PuO2 sources were made in the 2H container as a yellow II, Type A, special form 

material, with a 0.1 mR/hr Transport Index. 

5.3.1.3 Nickel Contamination Monitoring

Nickel-63 was used in the production of spark gaps. The radiation levels at the open ends of the gaps 

measured between 300 and 1,500 cpm. Losses of 63Ni occurred during vacuum firing and ultrasonic 

cleaning operations. Surveys of the process areas in the 888 Process Area showed no detectable 

63Ni (Jech 1963a, p. 7).

Nickel-63 was electroplated on a nickel mesh inside a sealed glass tube (a krypton). No internal 

monitoring was conducted. This is based on the types of operations with these devices, and routine 

survey data results of the work areas and equipment used. Fictitious worst-case doses were 

calculated and placed in an “Internal dosimetry technical notes file,” but this file has not been found 

(Weaver 1994a).

5.3.1.4 Carbon-14 Contamination Monitoring

There is no documentation on 14C monitoring even though it likely occurred.

5.3.2

5.3.2.1 Tritium Air Sampling

Air Sampling

Air-monitoring results were indicated in the 1962 (Jech 1963a) report for HTO and T2. A portable air 

monitor (T-290) was capable of an MDC of 1 × 10-4 μCi/ml, and a Kanne ion chamber area monitor 

was capable of a MDC of below 2 × 10-5 μCi/ml, the U.S. Atomic Energy Commission (AEC) 40-hr 

level. Typical work area measurements were less than 2 × 10-5 μCi/ml. Incidents approached the 

1 × 10-2 μCi/ml level. Maintenance levels averaged 1 × 10-3 μCi/ml. These levels usually occurred 

during maintenance operations in gloveboxes or vacuum pumps. The maximum tritium air 

concentration during normal operations was 6 × 10-4 μCi/ml. (Jech 1963a).

Room monitors were located in all areas where there was a potential for release of tritium. In 1971 

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. The alarm was set at about four times the AEC limit 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. The MDC has been calculated to be as low as 1.3 × 10-5 μCi/ml (Ward 

1971). Areas monitored included Laboratories 108, 109, 157, 158, 182, and 132. Portable samplers 

using silica gel collection media, or silica gel stations were also set up in some areas. 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. In addition, portable tritium gas (T2) 

monitors (the T-290 mentioned above) were used (GE 1987; Ward 1973).

5.3.2.2 Plutonium Air Sampling

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

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

5.5-cm-diameter filters. The following areas were monitored (Burkhart and Richardson 1986):
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• The receiving room where newly arrived heat sources were unpacked

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

• The defect analysis glovebox area where final units were disassembled

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

shipment were stored

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

stacks

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

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

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

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

sources on a monthly basis. 

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

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

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

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

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

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

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

plutonium (Eberline 1970). The Pinellas Plant continued to upgrade the system over the years (Dixon 

1988a,b). In the 1988 plutonium bioassay report it was indicated that no plutonium had been found 

detected in the air, on heat sources or on completed RTG units. (Weaver, 1989)

5.4 URINALYSIS VOLUME

Beginning in the 1960s HTO urine samples were submitted on a daily basis, at the end of the worked 

shifts, or at the beginning of the next shift. Spot samples were composited from each shifts’ individual 

samples to be counted on the next shift. If the samples were weekly, daily samples were composited 

to give the weekly sample that was then sampled and counted. 

For plutonium bioassays the sample size was what was received up to 800 ml. If the sample given 

was less than 800 ml, then the entire sample was analyzed. If more than 800 ml was submitted, only 

800 ml was used. Sample volumes were found to be documented for only several years. Because no 

dose assessments were made, the concentration values given in the analysis were likely based on 

the volume used in the analysis. Because volume information data was only found for several years, 

volume correction is not possible in most cases (Weaver 1989). 

5.5 ABSORPTION TYPES FOR RADIONUCLIDES INCLUDED IN THE BIOASSAY 

PROGRAM

The three radionuclides (T, 238Pu, and 239Pu) were handled in quantities or forms sufficient to present a 

concern for internal exposure according to an extensive review (Burkhart 1995a, p. 1). Three 

chemical forms of T were present: Water (HTO), metal hydride (primarily ScT2, ErT2 and TiT2), and 
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 16 of 31

gas (T2 ). Plutonium oxide (either 238Pu or 239Pu) was present between 1976 and 1991 only in triply 

encapsulated heat sources (Burkhart 1995a, p. 1).

ICRP Publication 68 lists plutonium oxides as absorption type S (ICRP 1995, p. 83). A discussion of 

absorption type for plutonium oxides in ICRP 71 indicates that bioassay data from accidentally 

exposed workers to 238PuO2 could have been closer to type M (ICRP 1996, p. 329). 

5.6 RADIONUCLIDES NOT INCLUDED IN THE BIOASSAY PROGRAM AND METAL

TRITIDES

Some Pinellas Plant health physicists investigated whether T metals, primarily titanium alloys, needed 

to be analyzed separately from HTO. 

The metal hydrides (primarily ScT2, ErT2, and TiT2) 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 remain intact (Burkhart 1995a p. 2). 

(Table 5-10 later in this document lists where incidents involving some of these materials occurred.) 

Intake prevention occurred by the elimination of the possibility of inhalation by means of ventilation 

and personnel protective equipment. There was little known of the biological behavior of these 

materials throughout most of the lifetime of the Pinellas Plant. The inability to detect the deposition of 

potentially small amounts of the low-energy beta-emitting tritium by whole-body counters or other 

external means would have made it very difficult to analyze for these soluble forms ICRP Publication 

71 indicates that one study in relation to TiT2 of an in vitro study using 1-μm particles (count median 

diameter) as a serum ultrafiltrate produced an assignment of absorption Type M (ICRP 1996, p 32). 

An in vivo study also indicated a Type M dissolution rate. ICRP 71 assumes that the biokinetics for 

metal hydride particles follows that of HTO. The Pinellas Plant provided the referenced researchers 

with the TiT2 to conduct the research (Weaver 1992a). A recent Health Physics journal article (Zhou 

and Cheng, 2004) categorizes TiT2 as ranking between type M and type F. The dose reconstructor 

should consider TiT2 as type M based upon ICRP 71 and the recent article in the Health Physics 

journal. Besides incidents, metal tritide intakes could have occurred from any of the process areas 

listed in Table 5-8 for HTO processing as presented in section 5.12.

Organic 14C and as CO2 are considered Class SR-2 (ICRP 1995, p. 22). The chemical form of 14C 

used at Pinellas is not known. The period of use was likely between 1979 and 1983, as evidenced in 

the environmental stack monitoring reports of that period (DOE 1979 – 1983).

The form of 63Ni is also not known, though it was plated on the vacuum switch tubes.

5.7 IN VIVO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PRACTICES

There was no in vivo monitoring at Pinellas.

5.8 INTERFERENCE AND UNCERTAINTY

5.8.1

“Uncertainties or errors for HTO 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 

Uncertainty and the Contamination of Samples
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 17 of 31

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

in Pinellas procedures (GE PDN 240001174, p48)”There was a possibility of cross contamination of 

urine samples. 

In about 1985 it was stated that one positive sample followed by a negative sample would negate the 

first positive sample if given in the same day the Pinellas Plant if the urine sample for tritium was 

unattainable the same day, the employee was asked to give the sample within the first hour of being 

on site the next day. It is likely that most personnel did this, thus minimizing the possible crosscontamination issue. There should have been no issue of cross contamination with plutonium samples 

because they were taken annually and not likely after recent work with the plutonium sources 

Uncertainties or errors for Pu bioassay measurements or the actual bioassay results themselves were 

usually not stated in the personnel records or database. MDCs were calculated based upon a 95% 

confidence level and samples were counted to a 2-sigma error as indicated in Pinellas procedures

(Weaver 1989) . Note, there were no verified positive bioassay results or known Pu contamination 

incidents. 

5.9 ASSESSMENT OF INTAKE FOR MONITORED AND UNMONITORED WORKERS

5.9.1

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 HTO and 

Plutonium potentially exposed workers have likely been monitored. Some conservative assumptions 

that a dose reconstructor should make are as follows:

Monitored Workers with Measurable Intakes (HTO and Plutonium)

• Radionuclide, particle size, and solubility type – These are listed below for bioassay 

radionuclides. If the work location of the employee is not known, default assumptions for 

radionuclide, particle size, and solubility type are equal to those in the most closely related 

facility. If the dose reconstructor decides to calculate plutonium doses, a 5-μm AMAD particle 

size should be assumed. Both HTO and T2 are absorption type V and have no characteristic 

particle size (ICRP 1995). 

• There is no correlation between tritium and plutonium exposures. 

5.9.2

A list of job titles that were not likely to receive a MT exposure is listed in Table 5-4 below. 

Metal Tritide Exposures

The remainder of this section on metal tritide exposure is reserved until additional exposure data can 

be researched and developed.

Table 5-4. Job Titles NOT likely to receive MT Exposures.

a,b

Job Title Job Title Job Title Job Title

secretary key operator office services coordinator buyer-equip-plant services

resources planning manger vendor material 

coordinator

data process computer 

operator

insurance pension clerk

inventory control 

coordinator

programmer analyst buyer financial analyst capital
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 18 of 31

Job Title Job Title Job Title Job Title

cashier-accounting clerk payroll clerk salary admin clerk mgr personnel relations

supervisor-computer 

operator

stenographer secretary –temporary accounting clerk

records retention control general accountant area dispatcher account payable clerk

drafter purchasing expediter receptionist general accounts clerk

graphic services aid special mfg program plan 

and scheduling

typist/clerk purchasing clerk

manager cost accounting spec graphic design financial analyst-contracts supervisor personnel 

accounting

group leader drafting supervisor-disbursements spec-telephone services personnel relations clerk

employee benefits clerk spec-employee store purchasing clerk media support coordinator

spec general accounting groundskeeper personnel account clerk

a. Sources: GE no date, PDN 240001303.

5.10 TRITIUM DOSE DISTRIBUTION INFORMATION

The dose calculation was initially estimated by assuming an acute dose integrated to infinity for a 

standard man, which yielded a calculation of about 5 mrem/μCi/L using a 12-day biological half-life. 

This result was used until about 1987 (Holliday 1983) when it changed to about 4.21 mrem/μCi/L (see 

Section 5.10.1 for the calculation).

It appears that most of the time the recording level was the basis for the calculation and recording of 

the infinite dose calculation. For example, if “<20” was indicated in the dose column listed as 

“mrem(B)” (even if 4 months equaled <20, four times the recorded dose was still zero), or if “<” or 

“<0.67” was indicated in the concentration column (μCi/L), no dose calculation was performed or 

added into the personal dose record. With the advent of the bioassay dose calculating computer 

program in about 1986, all doses were calculated but only results above 0.1 μCi/L were counted and 

cumulative doses above 1 mrem for the year were recorded in personnel annual. 

A health physics report for 1962 compiled a summary of urine sample data from 1959-1962 (Jech 

1963a, p. 6); Table 5-5 lists this information. The excretion rates are categorized into five 1-μCi/L 

bands. The maximum excretion levels are given along with the maximum personal dose due to tritium 

uptake.

The maximum yearly tritium deposition of 6.7 μCi/L was found in a technician who worked on a 

glovebox line vacuum ion pump in 1962. The air concentration was 6.5 × 10-4 μCi/cm2 with the

technician wearing a fresh air mask and protective clothing. The dose was calculated to be 70 mrem. 

The yearly dose of 0.2 rem was incurred by a technician engaged in normal hood work. A 

concentration of 5.3 μCi/L and 15 positive samples of 1 to 3 μCi/L occurred. The cause of the 

concentrations was traced to the handling of aging tank oil with compromised containment (Jech 

1963a, p. 6).

Table 5-6 lists data in dose bands for 1986 to 1991. The highest dose is recorded in parentheses. As 

shown, most of the monitored tritium dose was at values below 100 mrem/year during this period. 

Table 5-7 lists maximum and average tritium dose for the years that the information is available. This 

table will be updated as more information becomes available. The results discussed in Tables 5-5 and 

5-7 (up to 1975) utilized a 1.7 RBE for HTO. The results discussed in Table 5-6 utilized a RBE of 1.0.

Table 5-5. Summary of tritium urinalysis from 1959 to 1962.

a,b
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Description 1959 1960 1961 1962

Number of samples 506 657 712 836

Percent of samples <1 μCi/L 14.8 19.0 9.6 4.9

Number of samples 1 – 2 μCi/L 54 97 61 24

Number of samples 2 – 3 μCi/L 12 18 5 8

Number of samples 3 – 5 μCi/L 7 5 3 5

Number of samples >5 μCi/L 2 5 1 4

Maximum annual concentration (μCi/L) 7 7.2 7.4 6.7

Maximum yearly exposure (rem) 0.15 0.17 0.16 0.20

Number of persons 0.1 to 0.2 rem Not recorded Not recorded Not recorded 10

Number of persons 0 to 0.1 rem Not recorded Not recorded Not recorded 39

Number of persons sampled Not recorded Not recorded Not recorded 49

a. Jech (1963a, p. 6).

b. The dose calculations above included a 1.7 RBE for tritium.

Table 5-6. Summary of tritium urinalysis from 1986 to 1991.

a,b

Calendar 

year

Number of workers 

monitored

Number 

<10 mrem

Number 

<10–100 mrem

Number 

<101 – 500 mrem

Number 

>500 mrem

1986 194 39 21 (86) 0 0

1987 139 26 8 1 (105) 0

1988 129 28 13 1 (130) 0

1989 201 33 12 (97) 0 0

1990 177 17 6 (31) 0 0

1991 202 14 8 1 (101) 0

a. Source: Weaver (1992b).

b. A RBE of 1.0 was used.

Table 5-7. Summary of average and maximum tritium doses derived from tritium urinalysis by year.

b

Calendar 

year

Number of 

workers 

monitored

Average tritium 

dosea (mrem)

Maximum tritium 

dose (mrem) References

1956

1957

1958 367 GE 1958

1959 150 Jech 1963a

1960 170 Jech 1963a

1961 160 Jech 1963a

1962 49 200 Jech 1963a

1963-1975

1976 181 20.00 (7.29) 117 Holliday 1977

1977-1978

1979 236 Mauer 1979

1980-1982

1983 91 PDN 240001310

1984 236 PDN 240001310

1985 120 PDN 240001310 &1205

1986 194 86 PDN 240001205

1987 139 105 PDN 240001205

1988 129 16.85 (4.38) 130 PDN 240001205; Majestic 1989

1989 201 (2.77) 97 PDN 240001205

1990 177 7.21 (1.04) 31 PDN 240001205; Harder 1991

1991 202 12.76 (1.93) 101 PDN 240001205

1992 164 (0.91) 35 PDN 240001205
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Calendar 

year

Number of 

workers 

monitored

Average tritium 

dosea (mrem)

Maximum tritium 

dose (mrem) References

1993 134 (0.77) 21 PDN 240001205

1994 226 1.6 (0.80) 6 Weaver 1995

1995

1996

1997

a. ( ) Values indicate averages of positive and non-positive bioassays results averages. Non ( ) values indicate only 

positive bioassay result averages.

b. A RBE of 1.7 was used from 1954-1975. A RBE of 1.0 was used from 1976 –1994.

5.10.1

An RBE of 1.7 was used to calculate tritium exposures for years before 1976. This was suggested by 

the National Bureau of Standards (NBS), ICRP, and National Council on Radiation Protection and 

Measurements published references of that time. Exposures through 1975 were corrected to a 

1.0 RBE. Calculations for 1976 and subsequent years were made with an RBE of 1.0 (Holliday 

c. 1976). In the exposure records, most of the tritium dose records were corrected from 1957 to 1975 

by dividing by 1.7, as indicated in handwritten calculations in the claimants’ DOE dose files. All of the 

results discussed in Section 5.10 were not corrected for the 1.7 RBE. (Tables 5-5 and 5-7 used an 

RBE of 1.7, Table 5-6 used an RBE of 1.0). The infinite tritium dose calculations were used in the 

Pinellas computer program beginning about 1976.

Relative Biological Effectiveness for Tritium

The infinite tritium dose calculations were as follows (Weaver 1994b):

D# rem/μCi/day 

= (

(100 erg/g/rad) × (A or B) (5-3)

2.22 × 106 dpm/μCi) × (60 min/hr) × (24 hr/day) × (5.7 × 10-3 MeV/disintegration) × (1.6 × 10-6 erg/MeV)

The assumptions for these calculations were:

Total body water = 42,000 g A 

Total body soft tissue = 63,000 g B

Quality factor (rem/rad) = 1.0

Tritium effective half-life = 10 days

Tritium removal constant λ = 0.069/days

Using Assumption A, the value of D# rem/μCi/day is 6.94 × 10-6. Using Assumption B, that value is 

4.63 × 10-6. Therefore: 

D (infinity) = D#/λ × 42 L × 1,000 mrem/rem = 4.21 mrem/μCi/L (A) (5-4a)

D (infinity) = D#

/λ × 63 L × 1,000 mrem/rem = 4.21 mrem/μCi/L (B) (5-4b)

The tritium dose program used from 1987 calculated effective dose equivalent (EDE) using the default 

10-day effective half-life, or the program corrected the EDE up or down depending on the individual 

bioassay history (Weaver 1994b). A 5-mrem/μCi/L dose factor was used in the tritium program before 

1987, and 4.21-mrem/μCi/L was used from 1987 to 1997 (GE 1986).

The trigger for the GEDOSE computer program used since 1987 was 0.1 μCi/L, which for any single 

sample (if exceeded) led to calculation of dose from all sample concentrations above zero (Burkhart 

1995a). 
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The default effective half-life was 10 days, but if a different half-life was observed, that half-life was 

used in the calculations. 

5.11 IN VITRO URINALYSIS COUNTING FORM CODE INFORMATION

Because the claimant records sent by DOE contain handwritten cards that show all of the recorded 

internal dose information for a claimant, discussion of the various computer printouts generated by 

either Pinellas or a vendor (Landauer) is not necessary.

One records issue that could be confusing is the determination of the difference between the 

recording level for the MDC and the recording level for what was considered a positive dose. As 

illustrated in Table 5-2, there were a number of variations over the 40-year history of tritium 

monitoring. From 1957 through 1959 no tritium bioassay result was considered positive below 1 

μCi/L, even though the recording level for the bioassay result varied from <1 μCi/L down to 

0.01 μCi/L. In addition, nothing under a 5-mrem dose was considered positive. There are, however, 

secondary entries in the 1957 and 1958 records written next to the doses that were calculated as 

above 5 mrem which were counted. These values were used in the handwritten summary reports 

written in about 1985. (Some of these summaries failed to add some internal doses.) As of 1960 the 

recording level for tritium MDL was considered a less-than value, which was probably 0.67 μCi/L but 

the dose that was considered positive needed to be greater than 20 mrem. 

In comparing calculated doses to the actual recorded doses in the dosimetry records (1990 and 

1991), if the dose was calculated to 10.8 mrem, 10 mrem was entered. The result was rounded 

down. 

5.12 TRITIUM AND PLUTONIUM POTENTIAL RADIOACTIVE MATERIAL USE AREAS

There were a number of areas at the Pinellas Plant that had the potential for tritium or plutonium 

radioactive material area designations (see Tables 5-8 and 5-9). There have been a number of 

incidents with tritium contamination, as discussed in Section 5.13. There were no records found of 

plutonium incidents involving contamination. 

Table 5-8. Potential tritium radioactive material use areas. 

Location Name Activity

107 Tube Assembly Vacuum tube manufacturing & coating

108 Tube Exhaust and Test Vacuum tube evacuation and testing

109 Product Analysis Magnetic and radioactive gas leak checking

128 Tube Test

131 Final Tube Test

132M Fan Room Stack effluent control & tritium recovery

157/158 Gas Analysis Laboratory Hydrogen isotope analysis

182-C Tube Assembly Vacuum tube development and testing

182-G Tube Exhaust Vacuum tube development and testing

183 General Development

191 CPE Hood Room Ceramic product testing facility

Radioanalytical Laboratory Radiological Laboratory, tritium recovery

194 Engineering Environmental Testing

200 Test areas

800 Accelerator and Calibration

1000 Waste Storage

Source: DOE (1995, p. 5-3); Weaver (no date).
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 22 of 31

Table 5-9. Plutonium radioactive material 

management areas. 

Location Name

400 RTG Area

200 Test areas

800 Accelerator

1000 Waste Storage

5.13 INCIDENT HISTORY 

Table 5-10 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 HTO levels occurred in other areas. 

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

safety responses. On April 5, 1963, elevated amounts of HTO gas (up to about a level of 1 × 10-4 

μCi/ml) were noted on the room and stack monitors. Upon investigation the sniffer tube from the 

Kanne ion chamber was used to trace the contamination to a leaking shaft seal on the glovebox line 

recirculation pump. Corrective actions included increased hood ventilation and closure of the uranium 

tritide bed in Glovebox #1. All involved personnel, including those in health physics, were sampled for 

HTO. All results were below the 1-μCi/L MDC of that time (Jech 1963b).

On November 5, 1969, Cell #3 in Building 400 was found to be contaminated with HTO from stored 

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

but the air concentration of the exhaust measured about 600 μCi/m3. Bioassays of personnel in 

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

50 mrem (GE 1969). 

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

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

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

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

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

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

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

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

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

the release to Room 108 (Mauer 1979).

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

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

600,000 dpm/100 cm2

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

the vendor’s site and provided urinalysis for all requested personnel (Burkhart 1995b). 
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 23 of 31

Table 5-10. Chronology of unusual events and significant activities in relation to internal dosimetry. 

Date Description Curies released

of Tritium

12-57 Foot monitor installed in Area 108

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

2-11-58 Error estimating amount of tritium remaining in charging system in Room 18 1,253

7-8-58 Glass system breakage in Room 22 280

3-7-58 Glass system breakage in Room 18 567

8-16-58 Operator error with tritium loader valve position – Room 21 780

8-18-58 Glass manifold breakage – Room 21 1,180

2-10-59 Operator error in valve positioning – Room 8 286

2-20-59 Hand contamination – operator not wearing gloves

2-21-59 Area contamination – operator broke glass system

3-12-59 Operator contaminated during system cleaning by another worker (Room 14)

6-4-59 Personnel error working on SECS test system – Room 21 753

6-5-59 Area contamination-diffusion pump exploded in hood 14

6-18-59 Excess air released from tritium loading system 423

1-1960 Operator error – stopcock left open on tritium loading system 40

2-5-60 Glass manifold broke from strain 72

2-11-60 Operator error – stopcock left open on tritium loading system 308

3-25-60 Operator error caused exposure to 3 employees

5-14-60 Broken flask caused area contamination – Room 10

6-21-60 Ion gage exploded – room 16

7-8-60 Sample bulb dropped – room 23 6.8

7-13-60 Manifold shattered, exposing worker – room 23

8-12-60 Contamination spread TiH2 in area 108 from broken flask

9-10-60 Requirement for full anti-contamination clothing in Area 108 reduced to lab coats for normal production 

operations

11-14-60 Began using NBS Handbook 69 (NBS 1959) for maximum permissible concentrations (MPCs)

4-61 Area contamination from system breakage

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

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

9-1963 Employees found falsely identifying urine samples

4-5-63 Uncontrolled spread of tritium air contamination

3-30-65 Broken flask – room 9

3-20-65 Flask explosion – room 12

5-1966 SECS cold water removal problems 252

1-27-67 Glove box vacuum pump oil degassed 32

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

6-18-68 Acid cleaning explosion – Area 181

2-69 Leaking flange at absorption pump in Area 108 8 

2-69 Area contamination when pump exhaust lines were cut during hood removal – room 2 8 

11-5-69 Area contamination in building 400 associated with D-bed ? 

11-69 Building 400 cell #3 contaminated with TiH2 from used flask storage; all flasks moved to burning pad west 

of 400.

1-70 Area contamination/personnel exposure from flaking tube part in gas lab

2-70 Area contamination from pressurized sorb pump (air expansion) – room 2

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

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

3-12-71 Copper gasket uncovered in Room 18 hood – high internal dose 7.3

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

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

12-1-71 High internal exposure – room 18 hood work 1 

4-1972 Area contaminated from liquid discharge in area 182D 1.5

8-3-72 Leaking absorption pump 12

5-73 Area contaminated with ErH2 film in Areas 156, 157 and 158
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 24 of 31

Date Description Curies released

of Tritium

-11-73 Fire in boom box- building 200 0 

1-31-75 Improper valve closure on uranium bed 150

2-10-75 Absorption pump leak – area 182D 42

1-30-76 Contaminated 6 inch valve 0 

4-13-76 Oven fan blade broke tubes 0 

2-1977 Packaging fixtures in area 182D glove box 28

9-11-79 Work in Room 18 hood – internal dose 5.7

4-80 Area contaminated from film flaking – Area 158B 0 

8-80 Contaminated electron microscope 0 

4-81 Three waste drums found out gassing during truck loading were removed to Area 108; a procedure to

check all drums for out gassing prior to transfer to the storage building was set up. 

2-25-82 Tritium recovery system (TRS) valve left in wrong position after maintenance 8.6

4-20-82 Operator left TRS valve in wrong position after maintenance – area 108 48

5-24-82 TRS valve left in wrong position after maintenance – area 108, exhaust unit 513 9.5

9-1-82 Sample bulb leak 3 

1-5-83 Tritium storage bed oxidation problems 130

1-19-83 Absorption pump leak – area 208 9 

4-5-83 Bed heater control failure – Area 108 0 

5-84 Tritium air monitor system in Area 182 was rearranged

7-25-84 Absorption pump sieve dumped into drum in area 108 67

12-9-85 Sorb pump overheat – area contaminated 0 

6-24-86 Mass spectrometer oil change – workers exposed to tritium gas 1.5

11-4-87 Test of oxygen regeneration unit 12

2-11-88 Leaking sample bulb in area 108 8 

5-5-88 Purge left on over third shift in Area 108 – SECS overpressure 2.7

9-88 Lab area release over 2 week period 16.2

1-6-89 Water in SECS line vented in area 182D 1 

9-7-89 Loss of control of radioactive material 0 

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

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

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

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

often required bioassays after completion of the work (PDN 240001443; PDN 240001441). In 

addition, outside contractors were given HTO bioassays from the start of Pinellas operations in 1957.
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 25 of 31

REFERENCES

ANSI (American National Standards Institute), 1983, Internal Dosimetry Programs for Tritium 

Exposure – Minimum Requirements, N13.14-1983, New York.

Cited References

Burkhart, R. A., and J. A. Richardson, 1986, Building 400 Routine Air Sampling/Monitoring, June 20.

Burkhart, R. A., 1989, Release Criteria used at GEND, October 2 (PDN 240001010).

Burkhart, R. A., 1990, Report of Results of 1989 Plutonium HPQA Test Program, May 21.

Burkhart, R. A., 1995a, Internal Dosimetry Technical Basis Document, July 18, Lockheed Martin. 

(PDN 240001222)

Burkhart, R. A., 1995b, Status of Standard Leak Contamination Incident Investigation, May 23 (PDN 

240001452).

Christy, C. N., 1988, Environmental Plutonium Analysis, December 14.

Dixon, S. A., 1988a, Equipment Alpha Continuous Air Monitor, (Model alpha-3), March 24 (PDN 

240001343).

Dixon, S. A., 1988b, Equipment Alpha Continuous Air Monitor, (Model alpha-5), March 30 (PDN 

240001343).

DOE (U.S. Department of Energy), 1980, Environmental Monitoring Report 1979, Largo, Florida, April.

DOE (U.S. Department of Energy), 1981, Environmental Monitoring Report 1980, GEPP-EM-570, 

1597A/1598A/1610A-0080A, Largo, Florida, March.

DOE (U.S. Department of Energy), 1982, Environmental Monitoring Report 1981, GEPP-EM-654, 

8828A/8842A-0369A, Largo, Florida, March.

DOE (U.S. Department of Energy), 1983, Pinellas Plant Environmental Monitoring Report 1982,

GEPP-EM-729, 8828A-0369A, Largo, Florida, April.

DOE (U. S. Department of Energy), 1995, Pinellas Plant Environmental Assessment for 

Decontamination and Dismantlement, DOE/EA-1092, Albuquerque Operations Office, Albuquerque, 

New Mexico.

Eberline Instrument Corporation, 1970, Alpha air monitor model alpha - 1, May 1 (PDN 210001346).

GE (General Electric Company), 1958, Radiation Exposures – Outside Contractors (PDF 1256-1264).

GE (General Electric Company), 1964, Weekly Urinalysis Radiation Survey (PDN 240001042).

GE (General Electric Company), 1969, Tritium Building 400 1969, November 5 (PDN 240001329).

GE (General Electric Company), 1971, Liquid Sample Sheet (Urine Samples) (PDN 25000140)
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 26 of 31

GE (General Electric Company), 1973, Radioanalysis Sensitivities Table, November 8 (PDN 

240001349).

GE (General Electric Company), 1975, Visitor Tritium Bioassay Record, December 29 (PDN 

240000967).

GE (General Electric Company), 1983, Internal Dosimetry Practices, September 23 (PDN 

240001019).

GE (General Electric Company), 1984a, Visitor Tritium Bioassay Record (PDN 240000968).

GE (General Electric Company), 1984b, Assignment of Personnel to Work in Radioactive Material, 

Contamination or Radiation Areas, General Operating Procedure G.1.12, June 29 (PDN 240001264).

GE (General Electric Company), 1986, Tritium Dose Calculation Procedure, November 12 (PDN 

240001176).

GE (General Electric Company), 1987, Responses to Kanne Air Monitor Alarms, November 4 (PDN 

240001255).

GE (General Electric Company), 1989a, Health Physics Survey Schedule, February 10, Neutron 

Devices Department (PDN 240001195).

GE (General Electric Company), 1989b, MCA#1 Pu-238 & Pu-239 source, July 5 (PDN 240001343).

GE (General Electric Company), 1995, Health Physic Survey Schedule, October 2 (PDN 240001388).

GE (General Electric Company), no date, Radiation Badge Masters List with Job Titles, Neutron 

Device Department, St. Petersburg, Florida (PDN 240001303).

GE (General Electric Company), no date, Minimum Detectable Activity and Reporting Levels and 

Procedure for Processing and Counting Urine Sample in Beckman Counter, Neutron Device 

Department, St. Petersburg, Florida (PDN 240001174).

GE (General Electric Corporation), circa 1966, Smear Surveys Tritium, Neutron Device Department, 

St. Petersburg, Florida (PDN 240001177).

Harder, U. M., 1991, 1990 Annual ALARA Program Report for Ionizing Radiation, May 14 (PDN 

2400011202).

Holliday, J. N., 1977, Annual Summary of Whole Body Exposures to Ionizing Radiation, January 27 

(PDN 240001304).

Holliday, J. N., 1983, Internal Dosimetry Practices, September 23 (PDN 030000352).

Holliday, J. N., circa 1976, Tritium Exposures (PDN 030000352).

Huffman, G. N., 1979, Pu contaminated waste, January 30 (PDN 240001348)

ICRP (International Commission on Radiation Protection), 1994, Age-dependent Doses to Members 

of the Public from Intake of Radionuclides: Part 2 Ingestion Dose Coefficients, Publication 67, 

Pergamon Press, Oxford, England.
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 27 of 31

ICRP (International Commission on Radiation Protection), 1995, Dose Coefficients for Intakes of 

Radionuclides by Workers, Publication 68, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiation Protection), 1996, Age-Dependent Doses to Members 

of the Public from the Intake of Radionuclides: Part 4 Inhalation Dose Coefficients, Publication 71, 

Pergamon Press, Oxford, England.

Jech, J. J., 1963a, Health Physics Report - 1962, March 25 (PDN 240001044).

Jech, J. J., 1963b, Radiation Incident Investigation Report #3, April 5(PDN 240001044).

Majestic, J. R., 1989, 1988 Annual ALARA Program Report, March 10 (PDN 240001215).

Mauer, H. A., 1979, Investigation of Tritium Release and Personnel Exposure, October 11 (PDN 

240001326).

NBS (National Bureau of Standards), 1959, Maximum Permissible Body Burdens and Maximum 

Permissible Concentrations of Radionuclides in Air and Water for Occupational Exposure, NBS 

Handbook 69, Superintendent of Documents, Washington D.C.

PDN 240000887, Plutonium Bioassay Data (1975 through 1987), .

PDN 240000969, Individual Radiation Exposure Record, 1988 and 1989. 

PDN 240001043, Individual Radiation Exposure Record. 

PDN 240001051, Monthly Statistics and Summary 1960.

PDN 240001295, 1988 Pu Bioassay Results (1987 through 1990). 

PDN 240001441, Environmental Health and Safety Programs Work Permits, 1980-82.

PDN 240001443, Environmental Health and Safety Programs Work Permits, 1992.

Ward, W. E., 1971, Sensitivity Levels of Radiation Measurements, February 5.

Ward, W. E., 1973, Tritium Control Technology, July 16 (PDN 240001334).

Weaver, A. S., 1989, 1988 Plutonium Bioassay Results, January 27 (PDN 240000860).

Weaver, A. S., 1992a, “Support for Study by ITRI (Inhalation Toxicology Research Institute),” internal 

memorandum to T. Bryant (Martin Marietta Specialty Components), June 4.

Weaver, A. S., 1992b, 1991 ALARA Report, March 13 (PDN 240001203).

Weaver, A. S., 1994a, Internal Dosimetry Questions, November 22 (PDN 030000352).

Weaver, A. S., 1994b, Technical Specifications, November 24 (PDN 030000352).

Weaver, A. S., 1995, 1994 Annual ALARA Program Report for Ionizing Radiation, February 14 (PDN 

240001219).
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 28 of 31

Weaver, A. S., no date, Pinellas Plant Radiation Safety Overview and Visitor Orientation (PDN 

240001103). 

Zhou, Y. and Cheng,Y.S. 2004, “Dosimetry of Metal Tritide Particles as Evaluated by the ICRP 66 

Model and a Biokinetic Model from Laboratory Rats”, Health Physics Journal, vol 86, No. 2, February 

2004.

Author unknown, 1966, 1966 Bioassay Exposure Record. 

Other References

Author unknown, 1982, Air Monitor Checks, December 1 (PDN 240001179).

GE (General Electric Corporation), 1959, 1959 Urine Termination Records, Neutron Device 

Department, St. Petersburg, Florida.

GE (General Electric Corporation), 1990, Environmental Health and Safety, Historical Report on 

Radiation Protection at GE Neutron Devices, GEPP-EV-1126, Aerospace Division.

Hall, C. K., 1990, (1989) Annual ALARA Report for Ionizing Radiation, April 5 (PDN 240001214).

Holliday, J. N., 1976, Personnel Monitoring Reports – (monthly for 1976) (PDN 240001304).

Holliday, J. N., 1978, Update of RTG Safety Analysis Report Dated 12/15/1975, October 31 (PDN 

240001094).

Weaver, A. S., 1989, Air Sampling and Monitoring, February 10 (PDN 240001196)

Weaver, A. S., 1996, 1995 Annual ALARA Program Report for Ionizing Radiation, February 7 (PDN 

240001218).

Weaver, A. S., circa 1993, Internal Dosimetry Questions (PDN 030000352).
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 29 of 31

GLOSSARY

activity median aerodynamic diameter (AMAD)

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

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

acute

Pertaining to intakes received “acutely”, i.e., within a short period of time. 

bioassay

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

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

bioassay procedure

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

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

removed from the body.

body burden

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

time.

chronic

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

derived investigation level (DIL)

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

assay method.

dose 

A general term for absorbed dose, dose equivalent, effective dose equivalent, committed dose 

equivalent, committed effective dose equivalent, or total effective dose equivalent.

dose equivalent (H)

The product of absorbed dose (D) in rad (or gray) in tissue, a quality factor (Q), and other 

modifying factors (N). Dose equivalent is expressed in units of rem (or sievert) (1 rem = 0.01 

sievert).

exposure

The general condition of being subjected to ionizing radiation, such as by exposure to ionizing 

radiation from external sources or to ionizing radiation sources inside the body. In this 

document, exposure does not refer to the radiological physics concept of charge liberated per 

unit mass of air.

insoluble material

A term loosely used to describe the relative degree of solubility of a material in body fluids. 

Recognizing that no material is absolutely insoluble, the terms low solubility or poorly soluble 

are preferable.
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 30 of 31

intake

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

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

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

internal dose or exposure

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

sources).

internal dose assessment

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

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

In vitro measurement

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

in the excreta or in other biological materials removed from the body.

In vivo measurement

The measurement of radioactive material in the human body utilizing instrumentation that 

detects radiation emitted from the radioactive material in the body.

lung solubility type (F, M or S)

A classification scheme for inhaled material according to its rate of clearance from the 

pulmonary region of the lung.

minimum detectable amount

The smallest amount (activity or mass) of an analyte in a sample that will be detected with a 

probability, ∃, of non-detection (Type II error) while accepting a probability, ∀, of erroneously 

deciding that a positive (non-zero) quantity of analyte is present in an appropriate blank 

sample (Type I error).

minimum detectable concentration (MDC)

The minimum detectable amount expressed in units of concentration.

monitoring (personnel)

The measurement of radioactivity in the whole body, in a region of the body, in material 

eliminated from the body or in the air for reasons related to the estimation of intake of 

radioactive material. The term monitoring includes interpretation of the measurements.

occupational dose

An individual's ionizing radiation dose (external and internal) resulting from that individual's 

work assignment. Occupational dose does not include doses received as a medical patient or 

doses resulting from background radiation or participation as a subject in medical research 

programs.

radiation

Ionizing radiation: alpha particles, beta particles, gamma rays, X-rays, neutrons, high-speed 

electrons, high-speed protons, and other particles capable of producing ions. Radiation, as 

used in this part, does not include nonionizing radiation, such as radio- or microwaves, or 

visible, infrared, or ultraviolet light.
Document No. ORAUT-TKBS-0029-5 Revision No. 00 Effective Date: 10/07/2005 Page 31 of 31

recording level

A value below which data or results were considered to be too low to record and thus may not 

have been maintained.

rem

A special unit for dose equivalent. One rem is equal to 0.01 sievert.

routine monitoring

Monitoring carried out at regular intervals during normal operations.

sievert

The special name for the International System unit of dose equivalent. One sievert equals 1 

joule per kilogram, which equals 100 rem.

special monitoring

Monitoring carried out in actual or suspected abnormal conditions (i.e., measurements 

performed to estimate the amount of radionuclide deposited in a person when an intake is 

known or is suspected to have occurred). 

spot sample

A single void of urine.