ORAU Team

Dose Reconstruction Project for NIOSH

Y-12 National Security Complex – Occupational Internal Dose

Document Number:

ORAUT-TKBS-0014-5 

Effective Date: 10/11/2005

Revision No.: 01 PC-1 

Controlled Copy No.: ______

Page 1 of 44

Subject Experts: Bryce L. Rich and Melton H. Chew

Document Owner

Approval: Signature on File Date: 04/19/2005

William E. Murray, TBD Team Leader

Approval: Signature on File Date: 04/20/2005

Judson L. Kenoyer, Task 3 Manager

Concurrence: Signature on File Date: 04/20/2005

Richard E. Toohey, Project Director

Approval: Signature on File Date: 05/10/2005

James W. Neton, Associate Director for Science

Supersedes:

Revision No.: 00

TABLE OF CONTENTS

Section

Record of Issue/Revisions ...................................................................................................................

Page

4

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

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

5.1.1 Uranium Solubility in the Lung............................................................................................8

5.1.2 Particle Size Information...................................................................................................11

5.1.3 Uranium Alpha Activity as a Function of Enrichment ........................................................11

5.1.4 Temporal Pattern of Uranium Exposures..........................................................................13

5.1.5 Other Radionuclides of Concern Including Recycled Uranium 

Contaminants...................................................................................................................13

5.1.5.1 Unmonitored Radionuclides from Recycled Uranium............................................... 14

5.2 In Vitro Minimum Detectable Activities, Counting Methods, and Reporting Protocols................ 19

5.2.1 In Vitro Urine Analysis ......................................................................................................19

5.2.1.1 Coverage................................................................................................................. 19

5.2.1.2 Sample Collection.................................................................................................... 20

5.2.1.3 Minimum Detectable Activities ................................................................................. 21

5.2.2 In Vitro Methods for Individual Radionuclides ...................................................................23

5.2.2.1 In Vitro Bioassay for Uranium .................................................................................. 23

5.2.2.2 In Vitro Bioassay for Plutonium................................................................................ 25

5.2.2.3 In Vitro Bioassay for Tritium..................................................................................... 26

5.2.2.4 In Vitro Bioassay for Other Radionuclides................................................................ 26

5.2.3 Fecal Sample Analysis .....................................................................................................26

5.2.4 Interferences ....................................................................................................................27
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5.3 In Vivo Minimum Detectable Activities, Counting Methods, and Reporting Practices ................ 29

5.3.1 Whole-Body Counting ......................................................................................................29

5.3.2 Chest Counting ................................................................................................................29

5.3.2.1 Chest Counting for Uranium .................................................................................... 29

5.3.2.2 Chest Counting for Other Actinides.......................................................................... 30

5.3.2.3 Chest Counting for Other Radionuclides.................................................................. 31

5.4 Unmonitored Dose (NOT COMPLETE FOR ALL OPERATIONS AND/OR 

RADIONUCLIDES) ...................................................................................................................31

References ........................................................................................................................................32

Glossary ............................................................................................................................................36

Attachment 5A Occupational Internal Dose for Monitored Workers............................................... 39

5A.1 Conventions Used in Internal Dose Reports.............................................................................. 40

5A.1.1 Y-12 Urinalysis Data through September 1989................................................................. 40

5A.1.2 Y-12 Urinalysis Data since October 1989 .........................................................................40

5A.1.3 In Vivo Data......................................................................................................................41

5A.2 Uranium Solubility in the Lung................................................................................................... 43

5A.3 In Vitro Detection Limits ............................................................................................................ 43

5A.4 In Vivo Detection Limits............................................................................................................. 43
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LIST OF TABLES

Table

5-1 Enrichment assumptions for dose assessment for various analytical 

techniques .............................................................................................................................

Page

12

5-2 Y-12 Uranium receipts summary –presenting average levels of three 

predominant recycled uranium contaminants through plant history (1953 to 

1999)......................................................................................................................................15

5-3 Total receipts summary.......................................................................................................... 15

5-4 Primary contaminants (at 3-yr decay) in VHEU fuels and of concern in 

internal dose reconstruction ................................................................................................... 16

5-5 Comparison of the production of three selected isotopes in VHEU spent fuel 

with levels reported in HEU shipments to Y-12....................................................................... 16

5-6 Ranges of reported contaminant values in the main processing facilities 

(summary of Tables B-3 and B-4 of ) ..................................................................................... 17

5-7 Typical uranium enrichment materials at Y-12 with calculated and inferred 

RU contaminant levels ........................................................................................................... 18

5-8 Recommended defaults based on upper levels of expected ranges....................................... 19

5-9 LD values................................................................................................................................22

5-10 233U to 232U ratios ...................................................................................................................25

5-11 Reported lung counting detection limits for 235U ..................................................................... 30

5-12 Reported lung counting detection limits for 238U ..................................................................... 30

5A-1 Uranium urinalysis detection limits ......................................................................................... 43

5A-2 Other in vitro detection limits .................................................................................................. 43

5A-3 Reported in vivo detection limits for other radionuclides......................................................... 44

LIST OF FIGURES

Figure

5-1 Historical solubility types meeting the type M definition ............................................................

Page

9

5-2 Empirical and theoretical uranium excretion after uranium restriction..................................... 10

5-3 Lung deposition and retention of class W, class Q, and type M uranium................................ 12
Effective Date: 10/11/2005 Revision No. 01 PC-1 Document No. ORAUT-TKBS-0014-5 Page 4 of 44

RECORD OF ISSUE/REVISIONS

ISSUE 

AUTHORIZATION 

DATE

EFFECTIVE 

DATE REV. NO. DESCRIPTION

03/17/2004 03/17/2004 00 New document to establish TBD for the Y-12 

National Security Complex – Occupational 

Internal Dose. First approved issue. Initiated by 

William E. Murray.

05/10/2005 05/10/2005 01 First revision to include table numbers and titles 

in the body of the document. Approved issue of 

Revision 01. Initiated by William E. Murray.

05/10/2005 10/11/2005 01 PC-1 Page change initiated to incorporate the 

definition of U.S.C. on page 6 and details for the 

definition of a DOE facility on page 7. No 

sections were deleted. First approved page 

change revision for Revision 01. Retraining is 

not required. Initiated by William E. Murray.

Approval:

William E. Murray, TBD Team Leader

Signature on File______________10/06/2005 

Judson L. Kenoyer, Task 3 Manager

Signature on File _________ 10/04/2005 

Richard E. Toohey, Project Director

Signature on File________ 10/04/2005 

James W. Neton, Associate Director for Science

Signature on File_________ _ 10/11/2005 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 5 of 44

ACRONYMS AND ABBREVIATIONS

AMAD aerodynamic median activity diameter

Bq becquerel

CEDE committed effective dose equivalent

Ci curie

cpm counts per minute

d day

DF decontamination factor

dpm disintegrations per minute

DU depleted uranium (less than 0.72% 235U)

EU enriched uranium

g gram

hr hour

HEU highly enriched uranium (generally between 20% and 90% 235U)

HPGe high-purity germanium (detector)

ICRP International Commission on Radiological Protection

in inches

keV kilovolt-electron, 1,000 electron volts

kg kilogram

L liter

LC critical level (commonly 5% chance of a false positive)

LD detection level (commonly 5% chance of a false negative)

LEU low enriched uranium (generally less than 20% 235U)

MDA minimum detectable activity

mg milligram

min minute

mL milliliter

mm millimeter

MPBB maximum permissible body burden

MPLB maximum permissible lung burden

mrad millirad

mrem millirem

nCi nanocurie

NIOSH National Institute for Occupational Safety and Health

NU natural uranium

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory
Effective Date: 10/11/2005 Revision No. 01 PC-1 Document No. ORAUT-TKBS-0014-5 Page 6 of 44

pCi picocurie

ppb parts per billion

ppm parts per million

RWP Radiation Work Permit

RU recycled uranium

SMC Idaho Specific Manufacturing Capability

U.S.C. United States Code

VHEU very highly enriched uranium (greater than 90% 235U)

wt % weight percent

yr year

μCi microcurie

μg microgram

μm micrometer
Effective Date: 10/11/2005 Revision No. 01 PC-1 Document No. ORAUT-TKBS-0014-5 Page 7 of 44

5.1 INTRODUCTION

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 

(NIOSH) in the completion of the individual work required for each dose reconstruction.

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

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

employer facility” or a “Department of Energy facility” as defined in the Energy Employee 

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

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

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

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

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

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

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

facility encompassed by EEOICPA.

For DOE employees with cancer, the DOE facility definition only determines eligibility for a dose 

reconstruction, which is a prerequisite to a compensation decision (except for members of the Special 

Exposure Cohort). The compensation decision for cancer claimants is based on a section of the 

statute entitled “Exposure in the Performance of Duty.” That provision (42 U.S.C. § 7384n(b)) says 

that an individual with cancer “shall be determined to have sustained that cancer in the performance 

of duty for purposes of the compensation program if, and only if, the cancer ... was at least as likely 

as not related to employment at the facility [where the employee worked], as determined in 

accordance with the [probability of causation] guidelines established under subsection (c) ....” 42 

U.S.C. § 7384n(b). Neither the statute nor the probability of causation guidelines (nor the dose 

reconstruction regulation) define “performance of duty” for DOE employees with a covered cancer or 

restrict the “duty” to nuclear weapons work.

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

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

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

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

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

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

exclusion. Therefore, the statute requires NIOSH to include all radiation exposures in its dose 

reconstructions for employees at DOE facilities, including radiation exposures related to the Naval 

Nuclear Propulsion Program. As a result, all internal and external dosimetry results are considered 

valid for use in dose reconstruction. No efforts are made to determine the eligibility of any fraction of 

total measured exposure for inclusion in dose reconstruction.

At the Y-12 Plant, now the Y-12 National Security Complex, uranium isotopes in various chemical and 

physical forms have been the primary contributors to internal radiation doses to workers since 

November of 1943. Y-12 missions involved other radionuclides and, at times, some uranium 

compounds could have contained impurities with radiological health implications. These implications

and default analyses are discussed in this report. However, the primary focus on internal dose control 

has been on uranium compounds and alloys over a wide range of 235U enrichment. Therefore, this 
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section begins with an overview of concepts that apply broadly to the history of uranium work at Y-12. 

Attachment 5A summarizes some of the key points from the following sections.

There were programs in the early years of Y-12 operations for which the currently available 

information is insufficient to provide general guidance for internal dose reconstruction. One program 

was associated with the research engineering and development (RED) group, using a 

proton-accelerating cyclotron in Building 9201-2, which became operational on November 11, 1950 

and operated until 1980. At the peak of U-235 enrichment operations in the 1943-45 time period there 

were 1156 “large mass spectrographs”, called “calutrons” in operation. This enrichment production 

technique was shut down in 1946, although the calutrons in bldg 9204-3 were left operational for R&D 

in improving the production of the calutron concept and separating other element isotopes. Eight 

calutrons were still operating as late as 1997. This program produced an inventory of 225 isotopes 

from nearly every element on the periodic chart. The 86 inch Cyclotron started operation in 1950 and 

continued to operated until 1980. This cyclotron is no longer in operation. Polonium isotopes and 

alpha airborne activity are the mentioned internal dose concerns. The 86 inch cyclotron was later 

used to create neutron-deficient radionuclides as a part of the R&D program above. There were 

programs in these R&D efforts, which involved plutonium, and workers could have plutonium bioassay 

results in their records. When claim information indicates that a Y-12 worker was involved with 

research activities involving the Calutron, cyclotron (accelerator), fusion work, or plutonium (except in 

the case of recycled uranium (RU) exposure, which is addressed in this section), consideration must 

be given to possible exposure to radionuclides other than uranium. (Wilcox, Wm. J. Jr., et al. 1999)

5.1.1

The uranium compounds with which Y-12 has worked ranges from highly soluble to very insoluble. 

Health physicists have long recognized that nephrotoxicity is the primary hazard associated with 

soluble uranium compounds (depleted through low enrichment). According to Sterner and Riley 

(1946), exposures to soluble compounds were monitored from the closing days of World War II by 

clinical tests of renal function and by fluorometric tests for uranium in urine. However, no urinalysis 

data have been found for Y-12 prior to 1948, and this issue will have to be addressed in a revision of 

this document. 

Uranium Solubility in the Lung

Health physicists also recognized early that the lung is the primary organ of concern for the less 

soluble uranium compounds. During the greater part of the plant’s history, control measures for less 

soluble compounds were guided by the lung and metabolic models presented in the International 

Commission on Radiological Protection (ICRP) Publication 2 (ICRP 1960) or by similar predecessor 

models. These models related lung dose to uranium excretion in urine, particularly under chronic 

exposure conditions. 

It is noteworthy that insoluble material as described in ICRP Publication 2 would be classified as 

moderately soluble type M material in the modern ICRP Publication 66 framework (ICRP 1994). It is 

also noteworthy that ICRP has recognized that some forms of uranium are less soluble than initially 

believed. Figure 5-1 shows normalized lung retention for ICRP Publication 2 insoluble material, ICRP 

Publication 30 (ICRP 1979) class W particles, class Q material as described in the 1990s (Barber and 

Forrest 1995), and the Publication 66 type M particles. The boxes on the diagram define the modern 

clearance types F (fast), M (moderate), and S (slow) as a function of lung retention and time as 

illustrated in Annex D of ICRP Publication 71 (ICRP 1995). The ICRP 2 insoluble model and the class 

W models fit within the range of type M material. 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 9 of 44

0

0.2

0.4

0.6

0.8

1

0 30 60 90 120 150 180 210 240 270 300 330 360 390 420

Days

Fraction of Initial Alveolar Deposi

ICRP 2 (I)

Class'Q' (8um)

Type M (5 um)

Class W (1 um)

Type S

Type M

F

Figure 5-1. Historical solubility types meeting the type M definition. 

After in vivo lung counting was begun at Y-12 in 1961, it was recognized that a few workers had 

uranium lung burdens in considerable excess of the amount indicated by urinalysis. Writing in 1969, 

West and Scott (1969) noted that, “5 exposure cases noted in 1962 and 1963 showed chest burdens 

of enriched uranium [EU] which exhibited unexpectedly slow decreases after the persons were 

assigned to non-uranium activities” (West and Scott 1969). The job description of all five individuals 

was Process Operator (Snapp 2003). 

Subsequent investigations suggested that three process combustion ashes from the EU foundry were 

the most likely exposure materials for these five cases (Steckel and West 1966, p. 31). The particular 

thermal histories and particle size distributions of these process ashes were considered to be 

important factors leading to the low solubility in simulated lung fluid (Steckel and West 1966, p. 30). 

Particle sizes tended to be relatively larger (0.86 to 1.6 μm), and the solubility tended to be lower, for 

materials processed at higher temperatures.

While the exceptional cases with unusually protracted lung clearance are important, it is more 

important to note that, for the vast majority of individuals, lung clearance took place in approximate 

accordance with the ICRP Publication 2 (1960) insoluble model, which fits within the current type M 

framework. 

In 1964, a composite urinary excretion curve was described for 157 workers who had been removed 

from uranium work because of high uranium urinalysis results (Scott 1964). Figure 5-2 shows this 

empirical composite curve together with expected excretion curves for types F, M, and S uranium. 

The latter curves are derived from 5-μm aerodynamic median activity diameter (AMAD) excretion 

curves (Potter 2002). They were normalized to match the Scott curve at 200 d after removal from 

uranium work areas. The empirical excretion curve matches the type M curve much better than the 

type F or S curves. Indeed, the Scott curve fits the type M curve quite well over most of the 420-d 

time range. These results are consistent with the subsequent conclusions of Rucker et al. (2001) that 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 10 of 44

class W is the appropriate default for uranium at Bechtel Jacobs facilities, including Y-12. Exceptions 

to this default included high-fired uranium oxides and compounds formed by the slow oxidation of 

uranium metal.

0E+00

1E-05

2E-05

3E-05

4E-05

0 15 30 45 60 75 90 105 120 135 150 165 180 195

Days Post 300 Days Continuous Intake

Urinary Excretion (μCi/d)

Scott 1964

ICRP 68, F

ICRP 68, M

ICRP 68, S

Figure 5-2. Empirical and theoretical uranium excretion after uranium 

restriction. 

Beginning in September 1994, the plant was placed in a stand-down mode. This effectively stopped 

all routine work (and chronic exposure potential) in the process areas. Only minimal walkthroughs 

and area checks continued during the stand-down, which lasted until August 1998. The stand-down 

significantly influenced the available source term for exposures. Before stand-down, the Y-12 

program default modeling assumption was class Q (90% Super-W, 10% Y). During the stand-down, 

the Y-12 program default assumption of chronic exposure was modified to assume acute exposures 

occurring at the midpoint of a quarterly sampling frequency. No wet chemical operations were 

performed during this 4-yr period. Therefore, no soluble component was being produced and 

materials that were stopped in process were allowed to oxidize, which provided a larger component of 

type S material as an exposure source. In August 1998, a partial resumption of activities occurred; 

wet chemical operations were not restarted and were still not restarted as of October 2003. The 

predominant material encountered after partial restart in August 1998 is uranium oxide, which fecal 

sampling has shown to be more consistent with 100% type S material. Based on the changing 

workplace conditions after partial restart in 1998, the default assumptions were modified to return to 

chronic exposure but to use type S solubility. 

For a workplace as varied as Y-12, it is clear that no single solubility or particle size would apply to all 

workers. Furthermore, accurate assignment of the uranium lung clearance type to a given bioassay 

result was considered virtually impossible because of uncertainties about chemical form and 

limitations of the personnel-tracking system (Barber and Forest 1995, p. 669). As of 2001, Y-12 

implemented a radiation work permit (RWP) tracking system that tracks the location of workers, their 

type of work, and the type of bioassay needed (ORAU 2003, pp. 19 ff.).
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Exposure to type M material from 1948 to July 1998 appears to be the more likely absorption type. 

After July 1998, exposure to absorption type S material is more likely. However, the absorption type 

can be based on the monitoring data, claimant-favorable assumptions, or both. 

5.1.2

Several particle size studies in uranium process areas have been conducted at Y-12 (e.g., CCCC 

1953; Steckel and West 1966; Barber and Forest 1995; BWXT Y-12 2003a). For different times and 

different processes, reported particle sizes ranged from less than 1 to over 10 μm (physical). Steckel 

and West (1966) reported a positive correlation between uranium oxide particle size and process 

temperature. Barber and Forest (1995) used an 8-μm AMAD based on particle size measurements 

as the basis for the class Q dosimetry system used in the 1990s. (The class Q system is further 

described as 10% class Y and 90% modified class W. The modification consists of increasing the 

class W 50-d compartment to 120 d.) In terms of lung deposition and retention, the 8-μm AMAD class 

Q material is closer to 5-μm AMAD type M material than to 1-μm AMAD class W material, as shown in 

Figure 5-3.

Particle Size Information

Since 2000, Y-12 has implemented the latest guidance provided by the ICRP including the ICRP 

Publication 66 (ICRP 1994) lung model and ICRP Publication 78 (ICRP 1998) methods and models 

(BWXT Y-12 2003b, p. 89). This is the justification for the change to the 5-μm AMAD as the default 

particle size. Y-12 uses the newer weighting factors proposed by the ICRP in its 1990 

recommendations (ICRP 1991). For Y-12 dose reconstructions, the default particle size distribution is 

5 μm AMAD .

5.1.3

At Y-12, uranium enrichments range from depleted uranium (DU) [less than 0.72 wt % 235U] to very

highly enriched uranium (VHEU; greater than 93 wt % 235U). Section 5.1.4.1 lists the uranium isotopic 

mixes with recycled uranium (RU) contaminants that were typical of uranium materials at Y-12. The 

potential dosimetric complexity is simplified by the similarity of the dose conversion factors for the 

most important uranium isotopes and by the quantities measured in the in vitro bioassay and chestcounting programs. 

Uranium Alpha Activity as a Function of Enrichment

For work with enriched compounds and alloys, urinalysis programs after 1950 reported either total 

uranium alpha counts or isotopic uranium results, both of which reflect the dosimetric potential of the 

isotopes generally handled. Early in vivo chest-counting results were reported either as micrograms 

of 235U or milligrams of 238U based on whether an exposure was believed to be from general enriched 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 12 of 44

0.0001

0.001

0.01

0.1

1

1 10 100 1000

Days

Fraction of Intake in Lung .

Class W, 1 um

Type M, 5 um

Class Q, 8 um

Figure 5-3. Lung deposition and retention of class W, class Q, and type M 

uranium. (Particle sizes specified are in micrometers AMAD.)

uranium (EU), natural uranium (NU), or DU. Based on these reporting techniques, the following 

assumptions can be made about data interpretation for the periods from 1950 through 1989 for 

urinalysis and from 1961 to 1989 for lung counting. 

For lung counts a combination of the information in the Type Analysis and Material Type (see Section 

5.A.1.3 for details) reporting fields can be used to determine whether the count was believed to be 

due to NU or DU. For records through 1971, a Type Analysis of 1 with a Material Type of 2 or 3 

indicates DU, while a Material Type of 7 indicates NU. For records after 1971, a Type Analysis of 4 

with a Material Type of 2 or 3 indicates DU, while a Material Type of 7 indicates NU. 

Table 5-1. Enrichment assumptions for dose assessment for various analytical techniques. 

Analytical technique 

(Urinalysis 1948–present) 

(Lung count 1961–1989) Measurement information

Default enrichment 

assumption

Urine by fluorometry (usually for an 

individual who worked in areas with 

NU or DU; before 1950, also used 

for workers in EU areas.

Alpha dpm can be calculated from 

mass. 

Before 1950, 93% enrichment.

After 1949, NU. 

Urine by gross alpha counting 

(usually for an individual who worked 

in areas with EU).

Alpha dpm is the sum of all uranium 

alphas.

93% enrichment. 

Urine by alpha spectrometry

(10/1989 to present).

Alpha dpm/d is reported isotopically 

and total uranium is determined by 

summing isotopic results. 

Lung count μg U-235 93% enrichment.

Lung count μg U-238 NU.
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 13 of 44

At Y-12, it is current practice to assign dose by applying the dose conversion factor for 234U to all 

inferred uranium intakes. This practice is slightly conservative for EU; the resultant overestimate is 

less than 1% of the effective dose. For DU, the use of the 234U dose conversion factor leads to an 

overestimate of 16% for type S uranium and 20% for type M uranium (Eckerman and Kerr 1999, 

p. 16). 

5.1.4

As a rule, routine uranium exposures at Y-12 were considered to be of a chronic nature. For 

example, in 1958, Patterson stated, “...our interpretation of urinalysis results and our assignment of 

internal dose assume an exposure under equilibrium conditions of intake and elimination” (Patterson 

1958, p. 58). For the stand-down period from September 1994 until August 1998, acute exposures 

should be assumed to be the more likely mode of exposure. After this time, while complete 

equilibrium is not expected in modern internal dosimetry models, the presumption of chronic exposure 

conditions for uranium remains in place as noted in a 2003 report: “The most likely exposure potential 

for uranium work at the Y-12 Complex is chronic in nature” (BWXT Y-12 2003b, p. 107). Because 

workers were assigned to various buildings and processes throughout their employment, it is not 

practical to characterize intakes based on historical operational assignments. The claimant files do 

not generally contain detailed work assignments, but do contain all bioassay data for each individual 

worker during the years of operational exposure – primarily in vitro (urine) and in vivo (lung counts) 

bioassay. Radiological materials, including isotopic mixes and so forth, are not defined by operational 

area or building in this document, because internal dose reconstruction relies primarily on the 

recorded bioassay data and the defaults for unmonitored isotopes that were conservatively 

established on a plant-wide basis (see Table 5-8 later in this document for plant wide RU contaminant 

defaults).

Temporal Pattern of Uranium Exposures

5.1.5

In addition to 234U, 235U, and 238U, the following radionuclides were identified in the Y-12 Technical 

Basis Document for Internal Dosimetry (BWXT Y-12 2003a):

Other Radionuclides of Concern Including Recycled Uranium Contaminants

• 3H, 90Sr, 99Tc

• 228Th, 232Th

• 232U, 233U, 236U 

• 238Pu, 239Pu, 241Pu

• 237Np, 241Am

The following radionuclides that could interfere with in vivo analysis of uranium and thorium were 

quantified so that their effects on the spectra could be taken into account:

• 40K, 137Cs

Other radionuclides addressed included:

• 60Co and 95Zr/95Nb for organizations outside Y-12 (Cofield 1961)

• 210Po

Many of these radionuclides are accompanied by progeny in various stages of equilibrium. In 

addition, operators of the in vivo analysis equipment reviewed accumulating spectra for interferences 

from medical radioisotopes such as 131I.
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 14 of 44

5.1.5.1 Unmonitored Radionuclides from Recycled Uranium

One of the primary sources of significant contaminants (including the production of other than the

normal uranium isotopes) that were introduced into the Y-12 plant systems were those associated 

with RU from 1953 until 1999. These contaminants were a result of fission and activation processes 

of a variety of uranium enrichment isotope mixtures, which were irradiated in production and test 

reactors. After completing their useful life in the reactors, the unused uranium in the spent fuel 

elements or targets was recovered in chemical extraction plants and returned to the inventories in the 

DOE system along with trace quantities of the contaminants (in some instances U-236 was produced 

in quantities greater than “trace” levels).

The major source of RU contaminants at Y-12 was in the receipt and processing of returned VHEU 

from SRS and the Idaho Chemical Processing Plant (ICPP). Tables 5-2 and 5-3 document the 

quantities of uranium and the calculated mass quantities of plutonium, neptunium, and technetium 

actually delivered to Y-12. Table 5-4 presents a computer-generated listing of the relative production 

of a few isotopes of concern, following irradiation of VHEU fuel in a high-flux test reactor. The process 

streams included a spectrum of chemical forms of uranium, including uranyl nitrate [UO2(NO3)2] 

solutions, uranium trioxide (UO3), uranium-aluminum (U-Al) alloy ingots, uranium scrap, uranium 

tetrafluoride (UF4), uranium metal, solvent extraction raffinates, and a variety of secondary process 

wastes and residues.

The uranium enrichments ranged from DU, NU, low enriched uranium (LEU - less than 20% 235U), 

highly enriched uranium (HEU - from 20 to approximately 93% 235U), and very highly enriched (VHEU 

- more than 93% 235U). The predominant RU and associated contaminants were in HEU and VHEU 

materials. However, all of the uranium at Y-12 came from other DOE facilities, which had also either 

generated and/or received RU materials. Therefore, nearly all of the uranium in the DOE facilities 

contained RU contaminants to varying degrees through being processed in the same equipment, 

blending with other materials to adjust the degree of enrichment, etc. For example, most of the DU 

received and processed at Y-12 was seldom chemically processed but was received in forms from 

which parts were produced by mechanical processing. However, even these materials were received 

from other plants, such as Fernald, which also had an RU contaminant inventory. Therefore, the DU 

contaminant levels at Y-12 were inferred from both the Fernald and the Idaho Specific Manufacturing 

Capability (SMC) projects’ recycle reports (ORAUT 2004 and DOE 2000).

The fundamental conclusion from a variety of reports, including a lack of definitive RU contaminant

bioassay data, is that for dose reconstruction purposes a default level of RU contaminants should be 

derived and applied as a percentage increase to the derived uranium intake for each of the four major 

contaminants. Analytical information derived from a variety of sources allows the calculation or 

interpolation of the levels of the predominant RU contaminants in the uranium materials received, 

processed, and handled at Y-12. One of the primary sources of RU contaminant information was the 

documented report generated as a consequence of a comprehensive study of Recycled Uranium in 

DOE facilities in 2000. The applicable report for Y-12 was issued in December 2000 and titled 

Recycled Uranium Mass Balance Project Y-12 National Security Complex Site Report (DOE 2000). It 

should be noted that there were some recognized inconsistencies in shipping and receipts between 

the various DOE facilities in the DOE 2000 RU reports. In an attempt to resolve these inconsistencies 

a summary report in 2003 by the DOE Office of Security, titled Recycled Uranium, United States 

Production, Enrichment, and Utilization (DOE 2003) was issued though correcting some 

inconsistencies, there remain significant inconsistencies in the two reports. For purposes of 

reconstruction of RU contaminant inventories in the Y-12 facilities and internal dose default analyses, 

the rationale for resolving those inconsistencies is provided herein. One approach to gain an estimate 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 15 of 44

of the average contaminant concentrations through the years is to document the total receipts of 

contaminants and uranium and then to derive the average long-term RU contaminant concentrations. 

Tables 5-2 and 5-3 list the documented RU contaminants and their primary source in recycled VHEU. 

For completeness, other DU receipts are listed by the DOE (2000) report without documenting the 

associated RU contaminant levels. However, it is known that these uranium receipts also contained 

low-level RU contaminants, the concentrations of which are discussed later in this report. 

Table 5-2. Y-12 Uranium receipts summary –presenting average levels of three predominant recycled 

uranium contaminants through plant history (1953 to 1999). 

Site U Kg

Pu Np Tc

g ppb pCi/μg U g ppm pCi/μg U g ppm pCi/μg U

Y-12 VHEU Receipts Summary – DOE 2000 RU Report

SRS 125,161 0.0455 0.36 5.2E-03 3,600 29 2E-02 14,268 114 1.9

ICPP 25,696 0.00124 0.05 7.2E-04 66 3 2E-03 231 9 0.15

EU Total 150,857 0.04674 0.3 4.3E-03 3,666 23 1.6E-02 14,499 96 1.64

DU Receipts Summary - DOE 2000 RU Report

Hanford 1,502 No Data

ORGDP 192,836

PGDP 38,423

DU total 232,761

Grand total 383,618

a. ICPP = Idaho Chemical Processing Plant; ORGDP = Oak Ridge Gaseous Diffusion Plant; PGDP = Paducah Gaseous Diffusion Plant.

Table 5-3. Total recycled uranium receipts summary. 

Shipping site

U 

(kg)

Pu Np Tc

g ppb pCi/μg U g ppm pCi/μg U g ppm pCi/μg U

SRS 153,000

ICPP 25,700

Hanford 4,400

West Valley 900

Total 184,000 0.5 2.7 0.039 54 0.003 0.00021 9,100 49.5 0.85

The results of the DOE Office of Security RU report (DOE 2003) are presented in Table 5-3. As 

indicated there are significant differences with the results recorded in DOE (2000) report (summarized 

in Table 5-2) and 5-3. The objective of DOE (2003) was to resolve the differences in the reported 

shipment amounts from the recycling plants (the “primary sites”) with the reported receipts at the first 

or ”primary receiving sites.” Also the detail in DOE 2003 related to uranium enrichment and specific 

RU contaminant concentrations from each site were not provided. (Note the lack of detail in table 5-

3.) In examining DOE (2003), the shipments from these “primary receiving sites” (following 

processing and incorporation of recycled materials into their process streams) to what can be called 

“secondary” receiving sites were not documented. For example, Y-12 received uranium materials 

from gaseous diffusion plants, Fernald, and others, which also had RU contaminant inventories. Thus 

the discrepancies in the current values reported in DOE (2000) and DOE (2003) could result from 

documenting the “primary” shipment and receipts in DOE (2003), while DOE (2000) accounts for 

“secondary shipments” as well. However, where large differences occur, the most clamant-favorable 

value is used in deriving the defaults for dose reconstruction, which are generally the results in DOE 

(2000).

Though plutonium, neptunium, and technetium isotopes were analyzed and documented in all of the 

facilities as the primary isotopes of concern, there was a significant level of 228Th detected in nearly all 

of the facilities at Y-12 and in reported process stream analyses (DOE 2000). To establish the source 

of this isotope and provide a basis for including it in the RU contaminants to be considered in the 

internal dose reconstruction, Tables 5-4 and 5-5 are included to demonstrate the production of this 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 16 of 44

isotope in VHEU fuels and the relative/effective “enrichment” during chemical reprocessing. The 

values in Table 5-4 were derived for purposes of the DOE (2000) Recycled Uranium report at the 

ICPP. The ORIGIN2 computer program results illustrate the production of isotopes during use of 

VHEU fuels (D.R.Wenzel 2000 and D.R.Wenzel “specific tables” of isotopic production---“). The 

isotopes of interest for purposes of this illustration are both the uranium isotopes and the plutonium, 

neptunium, technetium, and thorium. Thorium-228 is a contaminant that became an isotope of 

concern at Y-12, which is the primary receiving site of recycled VHEU fuel materials.

Table 5-4. Primary contaminants (at 3-yr decay) in VHEU fuels and of concern in 

internal dose reconstruction. 

Isotope Half-life (yr)

Activity

(μCi/μg Total U)

Activity

(μCi/g Total U )

Percent weight

g/100 g Ua ppm U

Activity ratio

(isotope/U)

Tc-99 2.11E+05 3.8E-04 3.8E+02 2.2E+00 2.2E+04 24

Th-227 18.72 d 3.3E-11 3.3E-05 1.1E-13 1.1E-09

Th-228 1.913 1.6E-07 1.6E-01 1.9E-08 1.9E-04 0.01

Th-229 7.34E+03 1.4E-12 1.4E-06 6.7E-10 6.7E-06

Th-230 7.7E+04 1.9E-11 1.9E-05 9.6E-08 9.6E-04 1.2E-06

Th-231 25.52 hr 1.7E-06 1.7E+00 3.2E-10 3.2E-06 0.1

Th-232 1.41E+10 3.5E-15 3.5E-09 3.2E-06 3.2E-02 2.2E-10

Total 1.9

U-232 6.89E+01 2.7E-07 2.7E-01 1.3E-06 1.3E-02

U-233 1.59E+05 2.6E-10 2.6E-04 2.7E-06 2.7E-02

U-234 2.46E+05 6.6E-07 6.6E-01 1.0E+00 1.0E+04

U-235 7.04E+08 1.7E-06 1.7E-00 7.8E+01 7.8E+05

U-236 2.34E+07 1.3E-05 1.3E+01 2.0E+01 2.0E+05

U-237 6.75 d 3.6E-08 3.6E-02 4.4E-11 4.4E-11

U-238 4.47E+09 5.6E-09 5.6E-03 1.7E+00 1.7E+04

Total 15.7

Np-236 1.15E+05 9.4E-11 9.4E-05 7.2E-07 7.2E-03

Np-237 2.14E+06 9.1E-06 9.1E+00 1.3E+00 1.3E+04 0.6

Np-238 2.117 d 9.2E-10 9.2E-04 3.6E-13 3.6E-09

Np-239 2.355 4.3E-08 4.3E-02 1.9E-11 1.9E-07

Total 9.1

Pu-236 2.851 3.6E-06 3.6E+00 6.8E-07 6.8E-03 0.2

Pu-238 87.74 3.6E-02 3.6E+04 2.1E-01 2.1E+03 2.3E+03

Pu-239 2.41E+04 1.9E-05 1.9E+01 3.1E-02 3.1E+02 1.2

Pu-240 6.54E+03 1.5E-05 1.5E+01 6.4E-03 64 1.0

Pu-241 14.4 1.5E-03 1.5E+03 1.4E-03 14 1.0E+02

Pu-242 3.73E+05 8.8-E-09 8.8E-03 2.3E-04 2.3 5.6E-04

Total 4.6E+04

Am-241 4.32E+02 9.6E-06 9.6E+00 2.8E-04 2.8 0.6

a. There were basically three types of fuel elements with varying enrichments (both at the start and after burn-up). 

The typical after burn-up enrichments were 78.21% for aluminum-clad, 78.42% for zirconium-clad, and 51.9% for 

stainless-steel fuel elements. Values represent the 78% after burn-up enrichments as representative, with typical 

1%+ for U-234.

Table 5-5. Comparison of the production of three selected isotopes in VHEU spent fuel with levels 

reported in HEU shipments to Y-12. 

Isotope

μCi/g U product 

from SRS

μCi/g U product 

from ICPP

μCi/g U 

in high burn-up VHEU 

reactor fuel b 

Approximate Derived 

Decontamination 

Factor (DF)

Th-228 0.01 – 0.03 NAa 0.16 10

Np-237 0.01 – 0.02 1E-03 9.1 1E+03–1E+04

Pu-238 4E-03 3E-04 3.6E+04 1E+07–1E+08

a. There were no analytical data for Th-228 reported in the ICPP Y-12 receipt reports. The DF reported is interpolated 

from Y-12 process stream analytical data. These values were taken directly from column 4 in table 5-4 above.
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 17 of 44

Comparing the concentrations in the spent fuels with the reported concentrations in the product 

received at Y-12 provides a rough indication of the decontamination factors (DF) of the chemical 

extraction processes. Reference to these DFs in Table 5-5 and the average contaminant 

concentrations in Table 5-6 indicate that plutonium is the most efficiently removed with an effective DF

of approximately 1 x 107 to 108

, with a neptunium DF of approximately 1 × 104, while thorium appears 

to follow the chemistry of uranium more closely, resulting in a DF of only 10, which in turn results in a 

relative “enrichment” of 228Th in the VHEU product received at Y-12. Although 228Th has not been 

considered an RU contaminant of concern in other DOE facilities and was not routinely analyzed, 

there were sample analyses of contaminants in surface smears and wipes in most of the Y-12 plant 

facilities in 1997 and 2001, which included 228Th . These surface contamination surveys of a large 

number of Y-12 plant facilities indicate a predominance of 228Th as a contaminant. The uranium to 

non-uranium alpha ratios range from 4 to more than 1 × 104

, which is not useful for dose 

reconstruction. However, these data do indicate that 228Th is one of the primary contaminants at Y12, which in turn provides a technical base for inclusion of 228Th in the list of RU contaminant isotopes. 

Table 5-6. Ranges of reported contaminant values in the main processing 

facilities (summary of Tables B-3 and B-4 of ).

a 

Contaminant Range Ave. of range

Ave. of range 

minus the high

Raffinate 

valuesb

Pu 0.11 - 4.5 ppb 0.5 ppb 0.25 ppb 62 ppb

Np 4.7 – 346 ppm 30.6 ppm 20.8 ppm 2,980 ppm

Tc 0.13 – 211 ppm 83 ppm 66 ppm 641 ppm

a. The highest values reported in the processing facilities were in just one of 17 process 

streams - that being the solvent extraction purification of UN. Uranium levels are still 

significant in this stream, which indicates a marked enrichment of contaminants relative to 

U in this stream. 

b. The raffinate streams are depleted in uranium to a range of 5 ppm U levels in the sludge. 

Therefore the parts per million of the RU contaminants relative to U are not meaningful 

(i.e., total Pu – maximum 30 and average 21 pCi/g net sludge, Np – maximum 12 and 

average 8 pCi/g sludge, Tc – maximum 1.2E+4 and average 3.7E+3 pCi/g sludge).

Uranium has always been the dominant contributor to collective internal dose at Y-12. Monitoring for 

other radionuclides has been performed on a limited basis. Both the relatively small concentrations 

and the difficulty of analyses have contributed to the lack of data. The isotopic characterization of the 

large variety of uranium enrichments/contaminants is illustrated in Table 5-7, which is an average 

representation of six specific generalized enrichment types handled at Y-12. The uranium to nonuranium alpha ratios in Table 5-7 are in the same general range as those measured in surface 

contamination in the plant facilities after years of operation. 

From the values in Table 5-6, it is evident that the maximum of the ranges in the plant tend to run a 

factor of approximately 10 above the average of the range for both plutonium and neptunium, with 

technetium at a factor of 3. A factor of 10 is also applied to the thorium levels. The defaults in Table 

5-8 were chosen consistent with all of the information discussed above with the objective of 

representing the maximum values of the documented RU contaminant levels in the Y-12 process 

streams and thus represent claimant-favorable values, and which can apply on a plant–wide basis. 

The application of the default level should be based upon the indication by the bioassay as to the 

likelihood of VHEU/HEU or LEU/NU/DU. The appropriate default ratio should be chosen based on the 

235U or 238U indicated in the bioassay results. For each microgram or pCi of uranium intake calculated 

based on bioassay data in the claimant files, an additional intake of each of the four RU contaminants 

should be added.

From a practical and simplified approach, when the results in the claimant files are given in mass units 

of uranium per liter or lung count, column 3 values can be used to add the RU contaminant activity to 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 18 of 44

Table 5-7. Typical uranium enrichment materials at Y-12 with calculated and inferred RU contaminant 

levels. 

Uranium Isotope

Activity 

(Ci/g) Mass fraction

Activity 

(pCi/μg total U) RU Cont. Activity (pCi/μg total U)

High burn-up (~50%) recycled VHEU from SRS

233 9.66E-03 1.00E-04 9.660E-01 Pu-238 6.20E-03

234 6.22E-03 1.28E-02 7.962E+01 Pu-239 1.00E-04

235 2.16E-06 5.22E-01 1.128E+00 Np-237 2.00E-03

236 6.50E-05 2.92E-01 1.898E+01 Tc-99 1.90E+00

238 3.36E-07 1.73E-01 5.796E-02 Th-228 3.17E-02

Total 9.99E-01 1.007E+02 Total alpha 4.00E-02

dpm/μg total U 2.24E+02 α dpm/μg total U 8.87E-02

U α:Non U α 2.52E+03

Lower burn-up (~25%) recycled VHEU from ICPP

233 9.66E-03 3.000E-06 2.90E-02 Pu-238 2.00E-03

234 6.22E-03 1.000E-02 6.22E+01 Pu-239 3.30E-05

235 2.16E-06 7.800E-01 1.68E+00 Np-237 2.00E-03

236 6.50E-05 2.000E-01 1.30E+01 Tc-99 1.50E-01

238 3.36E-07 1.700E-02 7.69E-01 Th-228 5.00E-02

Total 1.007E+00 7.69E+01 Total α 5.40E-02

dpm/μg total U 1.71E+02 α dpm/μg total U 1.20E-01

U α:Non U α 1.4E+03

Weapons-grade Oralloy VHEU

233 9.66E-03 0.000E+00 0.00E+00 Pu-238 2.00E-02

234 6.22E-03 1.100E-02 6.84E-01 Pu-239 2.00E-02

235 2.16E-06 9.300E-01 2.01E+00 Np-237 6.00E-02

236 6.50E-05 7.500E-03 4.88E-01 Tc-99 5.00E-02

238 3.36E-07 6.000E-02 2.02E-02 Th-228 5.00E-02

Total 1.009E+00 7.09E+01 Total α 1.50E-01

dpm/μg total U 1.57E+02 α dpm/μg total U 3.33E-01

U α:Non U α 4.73E+02

Recycled NU

233 9.66E-03 0.000E+00 0.00E+00 Pu-238 0.00E+00

234 6.22E-03 5.400E-05 3.36E-01 Pu-239 3.00E-06

235 2.16E-06 7.200E-03 1.56E-02 Np-237 2.40E-05

236 6.50E-05 0.000E+00 0.00E+00 Tc-99 2.00E-04

238 3.36E-07 9.927E-01 3.34E-01 Th-228 3.00E-05

Total 6.85E-01 Total alpha 5.70E-05

dpm/μg total U 1.52E+00 α dpm/μg total U 1.27E-04

U α:Non U α 1.20E+04

Recycled LEU (2%) 

233 9.66E-03 0.000E+00 0.00E+00 Pu-238 0.00E+00

234 6.22E-03 2.000E-04 1.24E+00 Pu-239 2.00E-04

235 2.16E-06 2.000E-02 4.32E-02 Np-237 2.00E-04

236 6.50E-05 0.000E+00 0.00E+00 Tc-99 9.20E-02

238 3.36E-07 9.798E-01 3.29E-01 Th-228 3.00E-04

Total 1.62E+00 Total alpha 7.00E-04

dpm/μg total U 3.59E+00 α dpm/μg total U 1.55E-03

U α:Non U α 2.31E+03

Recycled DU

233 9.66E-03 0.000E+00 0.00E+00 Pu-238 0.00E+00

234 6.22E-03 1.000E-05 6.22E-02 Pu-239 3.70E-06

235 2.16E-06 2.000E-03 4.32E-03 Np-237 2.70E-05

236 6.50E-05 3.000E-06 1.95E-04 Tc-99 4.00E-04

238 3.36E-07 9.980E-01 3.35E-01 Th-228 4.00E-05

Total 4.02E-01 Total alpha 7.07E-05

dpm/μg total U 8.93E-01 α dpm/μg total U 1.57E-04

U α:Non U α 5.69E+03

a. Values derived from a number of sources and should be considered as largely average values.

b. The recycled DU, recycled NU, and recycled LEU values were derived from Fernald data because this site both supplied and received 

uranium to and from all the other DOE sites, including Y-12. In addition, the recycled DU information was verified by the analytical data 

reported by the Idaho SMC facility for DOE (2000). 

c. The weapons-grade Oralloy VHEU and other burn-up example VHEU were derived from Y-12 data and ICPP reports. 

d. The Oralloy RU contaminant data was presented in the reference reports in very general terms (i.e., <0.1 μCi/g U for total Np and Pu 

and <0.05 μCi/g for “Any other radionuclide”).
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 19 of 44

Table 5-8. Recommended defaults based on upper levels of expected ranges. 

Isotope Default levels in pCi/μg U Default levels in pCi/pCi total Uc

VHEU & Oralloy LEU, NU, & DU VHEU & Oralloy LEU, NU, & DU

Tc-99 4 0.3 4E-02 0.15

Th-228a 0.5 0.003 5E-03 1.5E-03

Np-237 0.6 0.003 6E-03 1.5E-03

Pu-238b 0.2 2E-03

Pu-239b 0.2 (Oralloy only) 0.003 2E-03 - Oralloy 1.5E-03

Total non U 1.3 0.01

Total U 75 2

a. The Th-228 default is justified in concept due to the demonstrated existence of general plant-wide 

contamination and further indicated by the demonstrated production in VHEU fuels, which in turn is 

further validated in the process stream analyses documented in DOE (2000).

b. VHEU was used in the metal form at Y-12 and called Oralloy. Plutonium analyses for Oralloy were 

not isotope specific. For Oralloy VHEU only, Pu-239 is the chosen isotope because the total 

quantities of recycled VHEU (in which Pu-238 was the predominant plutonium isotope) returned to 

the gaseous diffusion plants was miniscule compared to the LEU returns. 

c. Reference to Table 5-6 indicates activity in picocuries per gram of total U for VHEU ranges from 71 to 

101; 100 is used for simplicity. For the LEU, NU, and DU, the picocuries per gram of total U ranges 

from 0.9 to 1.6; 2 pCi/g total U is used for simplicity.

the intake. However, when the bioassay units are in uranium alpha activity per unit urine volume 

and/or U isotope activity units, the values in column 4 can be used. Again, Table 5-8 values are 

maximum values expected (and claimant-favorable), while average values are basically a factor of 10 

less than the values for plutonium, neptunium, and thorium and a factor of 3 less for technetium. If a 

best estimate dose reconstruction is needed, average values (a factor of 10 less than the values in 

Table 5-8 for plutonium, neptunium, and thorium and a factor of 3 less for technetium) should be 

used.

Reference to Table 5-8 with relationship to the plutonium contaminant, the recommended isotope Pu239 should be assumed for Oralloy only, since Oralloy used at Y-12 did not come directly from the 

primary RU generating sites, i.e. it was enriched at the GDP facilities in which Pu-239 was dominant 

compared to Pu-238.

It is important to note that the existence of air contamination and personnel exposures from 232Th 

operations is not addressed in this section. This section deals only with RU contaminants.

5.2 IN VITRO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PROTOCOLS

5.2.1

5.2.1.1 Coverage

In Vitro Urine Analysis

Uranium enrichment activities began in the fall of 1943. A uranium urinalysis program based on 

fluorometry was used in conjunction with medical examinations to monitor for kidney damage from 

exposure to soluble uranium compounds (Sterner and Riley 1946). The earliest fluorometry data was 

dated in 1948. In 1950 workers in production areas were placed on a uranium fluorometry urinalysis 

program for the purpose of estimating internal exposure. The program was expanded to include 

certain maintenance workers in 1954 (McLendon 1960). Y-12 currently requires workers with a 

potential for internal exposures in excess of 100 mrem/yr committed effective dose equivalent (CEDE) 

to participate in the bioassay program (BWXT Y-12 2003b, p. 23). Until September 1989, routine 

urinalysis focused on EU, DU, NU, tritium, and plutonium. Beginning in October 1989, uranium 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 20 of 44

results were no longer classified as EU, DU, or NU. Rather, uranium results were reported as isotopic 

results based on alpha spectrometry analysis. Analyses for other radionuclides were performed on an 

as-needed basis.

5.2.1.2 Sample Collection

Sample Volumes

For most of the plant’s history, the primary urine collection method was a spot sample submitted 

Monday morning before entering the work area. That is, routine samples were submitted after a 

minimum of a 48-hr absence from the work area. [In the July 1 to December 31, 1951, Health Physics 

Progress Report (UCC 1952, p. 25), it was stated that Friday evening samples would be discontinued 

in favor of Monday morning samples.] The fraction of the daily void volume was estimated on the 

basis of the time between the sample void and the previous void. This rate method of estimating daily 

void volumes was used explicitly in the calculation of the daily radionuclide excretion. 

For example, the following formula was given for calculating EU excretion by electrodeposition and 

gross alpha counting (UCC 1965, p. 25):

dpm/d = 8 × (AP ÷ Eff) × (Vol ÷ TI) (3.1)

where: 

AP = Average number of counts in 30 min on a plate (disk)

Eff = Uranium recovery efficiency (%)

Vol = Volume of total void (ml)

TI = Time between sample void and previous void (hr)

8 = A constant incorporating time, count, and volume constants, including the 20-ml 

electroplating aliquot volume. [There is no information in UCC (1965) on what the 

constant includes or what the daily excretion was.] 

The term (Vol ÷ TI) also appears in the corresponding equation for fluorometric determinations of NU 

and DU (UCC 1965, p. 27). The use of the rate method to estimate daily urinary excretion (and 

hence, radionuclide elimination) contributed to the uncertainty associated with any given 

measurement; and the corresponding detection level is discussed below.

Since 1989, routine samples have been collected over a 24-hr period, typically while the employee 

was on a scheduled break from the workplace. Many workers have elected to submit a simulated

24-hr sample. This sample is obtained by collecting the last void in the evening (before retiring), any 

urine excreted during the night, and the first void the following morning. This procedure is repeated 

for two consecutive nights (BWXT Y-12 2003b, p. 70).

Routine Urine Sample Frequency

Urine samples were collected monthly in 1950, and weekly collection for some employees began in 

1951. By 1963, health physics personnel were basing the frequency of participation in the urinalysis 

program for each department on the most recent urinalysis results of that department. The 1963 

edition of the Y-12 Radiation Safety Manual (UCC 1963a) stated that the frequency of participation 

schedules was reviewed monthly and adjusted semiannually to meet the following criterion: “Sample 

at the frequency necessary to assure, with at least 95% confidence, that 95% of the individuals in a 

department have a quarterly average below the plant action limit.” This criterion was used within the 

limitations of a maximum frequency of once per week and a minimum frequency of once per quarter. 

Since the late 1980s, most personnel have been on a quarterly frequency. 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 21 of 44

The current practice is to schedule samples based on the RWP usage rather than a default frequency. 

Only if an RWP was used will a person be scheduled for a sample. If the RWP specified a fecal 

sample, a paired set of urine and fecal samples is scheduled approximately 60 d after entry date. If 

the RWP specified urine only, a urine sample is scheduled approximately 90 d after entry date.

5.2.1.3 Minimum Detectable Activities

Normal and Depleted Uranium in Urine by Fluorometry (1945 to 1989) and Enriched Uranium 

(1948 to 1950)

In fluorometry, the visible radiation emitted from a doped, sodium-fluoride bead illuminated by 

ultraviolet radiation is measured. Poisson counting statistics used in nuclear particle counting do not 

apply to this procedure. The detection limits were historically determined by testing the performance 

of a particular configuration against standards of known content. Early Health Physics Progress 

Reports give the minimum detectable limit as 5 ppb, which is a minimum detectable activity of 7 μg/d 

for a nominal 1.4 L/d urine excretion rate (UCC 1949, 1950). In 1952, the instrument sensitivity and 

sample volumes indicated an MDA of 7 μg/d for 1.4 L/d of urine excreted (CCCC 1953, p. 28). 

The activity was calculated from the fluorometric mass reading for NU using a specific activity of 1.55 

dpm/μg. For EU (93 wt % 235U), a specific activity of 150 dpm/μg was used. The corresponding limits 

of detection are 11 dpm/d for NU and 1,100 dpm/d for EU (93 wt % 235U). At times, procedures called 

for the direct conversion of fluorometer current into units of disintegrations per minute, implicitly 

omitting the mass calculation. Whether the results were expressed in terms of mass or activity, the 

method remained essentially stable until replaced by long-duration alpha spectroscopy in 1989. 

Therefore, from 1945 through 1989, the MDA was 7 μg/d or 11 dpm/d for NU. From 1948 to 1950, the 

MDA was 7 μg/d or 1,100 dpm/d

Enriched Uranium in Urine by Gross Alpha Counting (1950 to September 1989)

for EU (93 wt % 235U). Additional variation resulted from individual 

urine excretion volumes. Given the limitations of the rate method of estimating daily urine volumes, 

uncertainty in the excretion volume is likely to contribute significantly to the uncertainty associated 

with the detection limit of a single measurement.

Determination of the MDAs for this method is complicated by the standard counting method used. 

From the mid-1950s to 1989, each sample consisted of two silver disks, each containing uranium 

electroplated from separate 20-ml raw urine aliquots drawn from the container submitted by the 

worker. Each disk was counted twice (on two different proportional counters) for 30 min per count. If 

the two results from a single disk did not agree within tabulated limits, a third count was made (UCC 

1966, p. 6) and the two most concordant counts were used. If the average results of the two disks 

from the same sample did not agree within specified limits, then two more plates were prepared, 

volume permitting (UCC 1966, p. 6). The expression of potentially censored data of this sort in terms 

of formal detection limits is not straightforward, and no detailed analysis of the statistics of this 

process has yet been found. However, an approximate indicator of the detection limit can be 

determined from published values, if issues about statistical independence are set aside.

The background count rate for the proportional counters was reported as 0.12 cpm in 1963 (UCC 

1963a, p. 38). Counters were not used if the background was greater than 5 counts in 30 min 

(Hamrick 1958, p. 7). For a disk counted for 120 min (in fact, two disks were counted for 60 min 

each), assuming a well-known background and alpha of 0.05, the detection level LD (Currie 1968) is 

0.13 cpm. For 1,400 ml/d urine output, a nominal 0.5 cpm/dpm counting efficiency, and an average 

uranium recovery of 40% as reported by Patterson (1958), the LD would be 46 dpm/d. With an 

increase in average uranium recovery to 73% (UCC 1965 p. 33), LD would be reduced to 25 dpm/d. 

In reviewing historical gross alpha urinalysis data, Barber and Forest (1995) report a decision level of 

approximately 20 dpm/d. Average uranium recoveries between 1958 and 1965 have not been 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 22 of 44

identified, nor is it clear that the statistical convention used by Barber and Forest was identical to that 

used here. It is provisionally assumed that LD was 46 dpm/d before 1965 and 25 dpm/d after 1965.

Uranium in Urine by Alpha Spectrometry (October 1989 to Present)

 

However, given the limitations of the rate method of estimating daily urine volumes, uncertainty in the 

excretion volume is expected to add substantially to the uncertainty associated with the detection limit 

of a single measurement. Because recoveries were based on batch rather than individual 

measurements, uncertainties in recovery would also add to the uncertainty of the detection limit of a 

single measurement. 

The present alpha spectrometry system employs large urine aliquots, chemical separation of uranium, 

sample- specific tracers to determine recovery, and long counting times on low-background detectors. 

The resultant LD is approximately 0.15 dpm/d

Tritium in Urine by Gas Counting 

(BWXT Y-12 2002, p. 15). The LD varies with the 

uranium recovery of a particular sample, and is reported with the sample result. 

Throughout most of the plant’s history (the earliest recorded data was from 1957), tritium was counted 

by reacting urine with calcium to evolve hydrogen gas, which was counted in an ion chamber. The 

MDA for this method has not been identified to date. However, the MDA was almost certainly far 

below the plant action level of 0.25 μCi/ml (8 × 108 dpm/d) (UCC 1957, p. 13). Results are reported in 

microcuries per milliliter. 

Tritium in Urine by Liquid Scintillation 

The current system is to place 5 ml of middle distillate from a urine sample in liquid scintillation fluid 

for counting. Assuming typical method efficiencies, the MDA is 2,000 dpm/d

Plutonium in Urine by Gross Alpha Counting

. 

The plutonium in urine procedure used before 1988 involved chemical separation from a 24-hr void or 

a simulated 24-hr sample before gross alpha counting. The detection limit has not yet been identified, 

but was certainly far below the corresponding limit for uranium in urine, which was based on 20-ml 

aliquots. If the counting method was identical (same count time of 60 min), chemical recovery (40% 

to 73%), and counting efficiency (0.5 cpm/dpm) to the early uranium-counting method, the MDA for a 

24-hr plutonium sample (less 40 ml) would range from about 0.7 to 1.3 dpm/sample.

Plutonium in Urine by Alpha Spectrometry (1988 to Present) 

The a priori LD for both 238Pu and for 239,240Pu is 0.025 dpm/sample, 

Other Actinides in Urine and Feces

or 0.025 dpm/d, assuming a full 

24-hr void volume. Actual sample-specific MDAs are included in the analysis reports.

Table 5-9 lists the analytical laboratory LD values used since about 1988 (BWXT Y-12 2003b, p. 107).

Table 5-9. LD values. 

Radionuclide LD (dpm/sample)

241Am 0.05

228Th 0.15

232Th 0.07

237Np 0.1

Uranium in Feces

Fecal sampling was started in 1998 when the presence of a less soluble component was identified as 

a result of the stand-down. In 2002, out of 2,800 participants in the bioassay program, 700 Y-12 

workers participated in the fecal-sampling program (BWXT Y-12 2002, p. 7). Participation in the 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 23 of 44

fecal-sampling program is based on the potential for significant exposures to insoluble uranium [i.e., 

workers who are expected to have an internal exposure to type S uranium and there is a potential to 

exceed 100 mrem (BWXT Y-12 2003b, p. 80)]. At present, the analytical chemistry division cites the 

value of 0.15 dpm/sample

5.2.2

as a representative MDA for fecal analysis (BWXT Y-12 2002, p. 15). 

Samples usually consist of a single void. The results in the claim files are assumed to be in units of 

activity per sample unless otherwise indicated (BWXT Y-12 2003c, p. 7). 

The following section discusses the in vitro methods for specific radionuclides in urine. (Section 5.2.3 

discusses analysis of fecal samples.)

In Vitro Methods for Individual Radionuclides

5.2.2.1 In Vitro Bioassay for Uranium

Overview

The in vitro bioassay program for uranium at Y-12 can be divided into four eras. First, from 1943 to 

1947, limited monitoring occurred, for which the data cannot be retrieved. Second, from 1948 to 

1950, fluorometric analyses of urine and blood occurred as part of general medical surveillance to 

prevent kidney damage from exposure to soluble uranium compounds. Third, from 1950 to 1989, 

collection of routine samples based on uranium exposure potential was initiated. During this third 

phase, fluorometric analyses were performed on samples submitted by workers in NU and DU areas. 

Electrodeposition of uranium, followed by gross alpha counting, was used for samples submitted by 

workers in EU areas. The primary goal for the EU analyses was to control lung doses from insoluble 

compounds under the assumption that such materials cleared from the lung with a 120-d half-life. 

During this lengthy era there were modifications in procedures, but the basic approach remained the 

same. Fecal sampling was used for some follow-up investigations. Fourth, after October 1989, 

routine 24-hr urine samples were collected. The uranium was extracted by chemical separation and 

ion exchange, and the extract was counted by alpha spectroscopy for an extended period (1,000 

min/sample). Since 2000, fecal samples have been submitted routinely for individuals working with 

largely insoluble forms of uranium.

Uranium Analysis by Fluorometry (1948 to 1989)

Uranium processing began in the fall of 1943. Y-12 technical reports describe the use of fluorometry 

for uranium detection as early as 1944 (Van Wazer and Reiss 1944). By 1946, fluorometric analyses 

for uranium in urine and blood were in general use to supplement clinical surveillance for soluble 

uranium exposures (Sterner and Riley 1946). Fluorometric uranium urinalysis data from the 

Tennessee Eastman Corporation era (1943 to 1947) were discussed in a 1981 mortality study of plant 

personnel (Polednak and Frome 1981). 

Early in the Union Carbide management of Y-12, the fluorometric method was reassessed. Extraction 

methods were used starting in 1949 and 1950 (CCCC 1949, p. 7; UCC 1950, p. 14). By the first half 

of 1952, a technique involving small (0.2-ml) aliquots of raw urine had been put in place (CCCC 1953, 

p. 28). However, after this change in technique was instituted the urinary uranium concentrations 

increased greatly as shown in Table 5-7. This raises the concern that the technique used before May 

of 1952 could have underestimated the urinary uranium concentrations. (See Section 5.4.2 for 

discussion of missed doses.)

Fluorometric urinalysis continued to be used for NU and DU until 1989, when alpha spectroscopy 

began to be used. Fluorometry yielded results in mass units (e.g., μg/L), but results were often 

converted to disintegrations per minute per day with the assumed specific activity of 1.55 dpm/μg for 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 24 of 44

NU. The fluorometric technique had an industry standard sensitivity of about 5 ppb (5 μg/L) (e.g.,

UCC 1949, p. 7).

Uranium Analysis by Electrodeposition and Alpha Counting (1950 to 1989) 

The primary objective of this procedure was to prevent EU lung burdens in excess of the prevailing 

limit (0.017 μCi) corresponding to an ICRP 2 (ICRP 1960) calculated lung dose of 15 rem/yr (using a 

quality factor of 10). Samples from workers in EU areas were analyzed by electrodeposition of 

uranium onto silver discs, which were then counted for gross alpha activity in a proportional counter. 

Development of the gross alpha counting method was reported to be underway in early 1949 (CCCC 

1949, p. 7). At first, samples were subjected to an acid digestion step before electrodeposition. In 

1951, methodological problems leading to underestimates of a factor of 2 were noted and corrected 

(UCC 1952, p. 53). The information available was not sufficient to determine if the records 

themselves were corrected. In the mid-1950s, the acid digestion step was discontinued altogether 

due to reported contamination problems, as well as for logistical considerations (Patterson 1958, 

p. 34). 

From the mid-1950s, two 20-ml aliquots of raw urine from each sample were placed in 

electrodeposition cells, which produced two discs per sample. Each disc was counted once on one 

proportional counter for 30 min and a second time on a different counter. The results were then 

averaged. The background count rate was reported as 0.12 cpm in 1963 (UCC 1963a, p. 38). The 

expected net count rate for a person excreting at the action level of 70 dpm/d was 0.2 cpm, given a 

urine output of 1.4 L/d, a nominal 0.5 dpm/cpm counting efficiency, and a uranium recovery of 40%. 

Because each disk was counted twice for 30 min each, the expected net count for the two counts on 

each disk under these conditions was 12 counts, over a background of 7.2 counts.

Recoveries were estimated on the basis of spiked samples that accompanied each counting run. The 

typical spike contained relatively little activity. Recovery rates from raw urine tended to be around 

40% in the late 1950s (Patterson 1958), but they improved over time: In 1965, the average recovery 

for electroplating was 73% (UCC 1965, p. 33). Method limitations are discussed in a number of 

reports, including Johnson et al. (1959) and UCNC (1959). Recovery rates were found to vary with 

pH, drying procedure, and uranium content, with rates decreasing for uranium concentrations below 

the plant action level. The precision of any one sample was acknowledged to be relatively low: 

Further, realizing the over-all lack of precision in any one sample result, we normally 

recommend restriction only on the basis of the 13-week or quarterly cumulative internal 

dose as indicated by from 2 to 13 samples. (Patterson 1958, p. 57)

The precision of an individual result in this type of procedure is poor and for this reason 

little significance is attached to individual results, particularly those below the “Plant 

Action Limit (PAL)” level. (UCC 1966).

These limitations notwithstanding, the salient point is that, when in vivo counting became routine in 

the early 1960s, very few additional workers were found to require restriction from uranium areas 

(e.g., UCC 1963b, p. 6). The in vivo monitoring frequency was determined by urinalysis results and 

ranged from monthly to once each 18 months (UCC 1963b, p. 1). Almost all of the restrictions that did 

take place would have occurred on the basis of the urinalysis program alone. In those cases in which 

restriction was based on in vivo analysis alone, lung retention times were often observed to be 

considerably longer than the assumed 120-d lung half-life on which the urinalysis program was based. 

In other words, the problem was largely due to the inapplicability of the lung model for some materials 

and individuals, rather than a failure of the urinalysis program. 
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Uranium-233 by Gross Alpha Counting

In early 1962, Y-12 undertook the fabrication of metallic 233U from a nitrate solution (233U uranyl 

nitrate) in Building 9205 (West and Roberts 1962). In recognition of the high specific activity of this 

material, and the potential for gamma radiation from the progeny of contaminant 232U, operations were 

doubly contained. Pilot runs were made with less hazardous materials to test the protective 

measures. As part of these precautions, eight workers submitted urine samples for analysis by 

electrodeposition and gross alpha counting as described above, with the same recoveries and 

detection efficiencies. Ratios of 233U to 232U varied with operation and time. Table 5-10 list ratios for 

various exposure scenarios. In addition to absorption types M and S, exposure to type F material is 

considered likely in this process. Uranium-232 has the larger dose conversion factor. 

Table 5-10. 233U to 232U ratios. 

Process

U-233:U-232 

activity ratio

Reduction of UF4 1.3

Casting of metal-crucible skull 15–32

Machining final uranium component 160

Due to uncertainty in the process, claimant-favorable assumptions should be made about solubility 

and uranium activity ratios.

Uranium Analysis by Alpha Spectrometry (1989 to Present)

Up to 2 L of urine, along with a 232U tracer, are treated with nitric acid (HNO3) and hydrogen peroxide 

(H2O2). The sample is wet-ashed, and the uranium is coprecipitated with calcium oxalate (CaC2O2). 

After dissolving the precipitate in hydrochloric acid (HCl), the uranium is further separated by ion 

exchange chromatography. The uranium is eluted from the column with a solution of dilute HCl to 

which titanous chloride (TiCl3) has been added to reduce actinides that could be in an elevated 

oxidation state. The final fraction of the eluate is first treated with ascorbic acid to reduce the 

presence of any ferric iron and next with hydrofluoric acid (HF). Then the uranium isotopes are 

coprecipitated on neodymium fluoride (NdF3). The NdF3 is caught on a 0.1-mm filter, rinsed, dried, 

and mounted on a planchet for alpha spectrometry (BWXT Y-12 2003b). Typical recoveries are about 

85%. Samples are typically counted for 16 hr on a passivated implanted planar silicon detector.

5.2.2.2 In Vitro Bioassay for Plutonium

Plutonium in Urine by Gross Alpha Counting

In 1957, certain workers in the Special Testing Department were tested for plutonium intake by 

urinalysis. Because large urine volumes were required, these employees submitted 24-hr samples 

through the use of take-home kits. The sampling frequency was monthly (UCC 1957, p. 38). After 

two 20-ml aliquots were removed for uranium analysis by electrodeposition, the balance of the sample 

was treated by chemical separation, precipitation, and evaporation onto a stainless-steel planchet for 

alpha counting. Health physics personnel converted the reported activity rate (disintegrations per 

minute per 24-hr void) to millirem per day to the critical organ.

Uranium, Plutonium, and Americium by Alpha Spectrometry (October 1989 to Present)

Up to 2 L of urine, along with 232U, 242Pu, and 243Am tracers, are treated with HNO3 and H2O2. The 

sample is wet ashed, and the uranium, plutonium, and americium are coprecipitated with CaC2O2. 

After dissolving the precipitate in HCl, the uranium, plutonium, and americium are further separated by 

the use of two ion exchange columns and an additional oxalate precipitation. The uranium, plutonium, 

and americium in the final fractions are coprecipitated with NdF3. The NdF3 is caught on a 0.1-mm 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 26 of 44

filter, rinsed, dried, and mounted on a planchet for alpha spectrometry (BWXT Y-12 2003b). The 

same procedure is used for isotopic thorium.

5.2.2.3 In Vitro Bioassay for Tritium

As of 1957, personnel engaged in processing materials with a potential for tritium contamination 

submitted three urine samples per month. Samples were submitted at the same stations used for 

uranium and plutonium samples. The samples were reacted to evolve hydrogen gas, which was then 

counted for beta activity in an ionization chamber. Results were reported in microcuries per milliliter. 

(UCC 1957, pp. 38–39)

Tritium is currently measured by liquid scintillation counting. An aliquot of the urine sample is distilled 

in a heating mantle, and the middle fraction of the distillate is collected. A 5-ml portion of the collected 

distillate is mixed with liquid scintillation cocktail. The beta activity of the tritium is then measured by 

liquid scintillation counting (BWXT Y-12 2003b). Exposure should be assumed to be due to HTO. 

5.2.2.4 In Vitro Bioassay for Other Radionuclides

The following alpha spectrometry procedures described in Section 3.2 of BWXT Y-12 (2003b) have 

been in use since 1989.

Neptunium-237 in Urine by Alpha Spectrometry

In this procedure, 239Np is used as an internal yield monitor. Americium-243, which is in equilibrium 

with its daughter, 239Np, is added during the sample preparation process. The separation time of 

239Np from its parent, 243Am, is recorded (so that the count can be corrected for decay), and the beta 

activity is measured to calculate the radiochemical yield. The tracer is added to the urine sample (or 

fecal material) along with concentrated HNO3 and H2O2, and the sample is wet ashed. Neptunium is 

coprecipitated with CaC2O2 and mixed hydroxides. After dissolving the precipitate in concentrated 

HNO3 and ashing in a muffle furnace, the oxidation state of neptunium is adjusted to the tetravalent 

state by using ferrous sulfamate as a reducing agent. Neptunium is separated from other actinides by 

the anion exchange method. Neptunium is eluted with 0.5-molar HCl solution containing TiCl3. The 

eluate is collected, and neptunium is coprecipitated on NdF3 and filtered on 0.1-mm filter paper. The 

NdF3 is rinsed, dried, and mounted on a planchet for beta counting and alpha spectrometry.

Isotopic Thorium (228Th, 230Th, and 232Th) by Alpha Spectrometry

Up to 2 L of urine along with a 229Th tracer is treated with HNO3 and H2O2. The sample is wet ashed, 

and the thorium is coprecipitated with CaC2O2. After dissolving the precipitate in HNO3 and mixing 

with methanol, the thorium is further separated by ion exchange chromatography. The thorium is 

eluted from the column with 8-molar HCl. Hydrofluoric acid (HF) is added, and the thorium isotopes 

are coprecipitated with NdF3. The NdF3 is caught on a 0.1-mm filter, rinsed, dried, and mounted on a 

planchet for alpha spectrometry. The same procedure is used for uranium, plutonium, and americium 

determinations. This method accommodates substantial quantities of common interferences, such as 

iron, aluminum, titanium, zirconium, and uranium. However, plutonium can present some 

interference. If plutonium is expected, an alternate method should be chosen (BWXT Y-12 2003b).

5.2.3

Fecal samples have long been used in follow-up investigations when urinalysis or in vivo

measurements indicate the likelihood of a substantial intake. West and Scott (1966) describe such an 

investigation in the 1960s in which fecal, urine, and in vivo measurements were used jointly to 

investigate clearance. Fecal sampling was discontinued several times in the past, but was reinstated 

Fecal Sample Analysis
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in 1998 due to changes in workplace exposure conditions. At first, fecal sampling was used for a 

limited number of workers in Building 9212 but was expanded to include Building 9215 in 1999 and to 

other areas of the plant in 2000 and 2001 (BWXT Y-12 2002, p. 6). 

If a work area has been determined to have predominantly insoluble airborne uranium, it is current 

practice for workers with a moderate to high exposure potential (CEDE greater than 100 mrem) to 

submit both urine and fecal samples. The relative elimination by urinary and fecal pathways is used 

to determine the solubility mixtures for each individual. For combined urine and fecal monitoring, the 

routine sampling interval is 53 d (BWXT Y-12 2002, p. 14) with 24-hr fecal collections preferred. 

When 24-hr fecal samples are not available, the results can be normalized using the ratio of the 

Reference Man excretion rate of 135 g/d to the mass of the submitted sample. The Y-12 practice was 

to routinely normalize results to Reference Man. These normalized results or actual individual daily 

excretion results can be used to reconstruct dose.

5.2.4

As investigation levels decreased over time, the contribution of natural background to worker uranium 

excretion assumed greater importance. Since 1989, Y-12 has corrected the measured uranium 

excretion values for background uranium interference when performing dose assessments. However, 

the uranium results reported in response to NIOSH requests for dose records have not been 

corrected for dietary uranium.

Interferences

Y-12 used the following methods to correct sample results reported after 1989 for dietary levels of 

uranium. The range of background uranium values expected to be observed in urine samples is 0.14 

to 0.57 dpm and the range of background uranium values expected to be observed in the fecal 

samples is 2.32 to 2.75 dpm (BWXT Y-12 2003b, p. 75). Therefore, at Y-12, the reference point is 

established at the midpoint of the quoted ranges of values for urine (0.35 dpm/d) and fecal samples 

(2.5 dpm/sample). The ratios between 234U and 238U can also provide information to help characterize 

the sample. Natural background uranium typically has a 234U to 238U ratio of approximately 1:1. EU 

also has a characteristic ratio, typically greater than 3:1. DU, on the other hand, typically has a 

characteristic ratio of less than 0.9:1.

Based on the above information, the background uranium corrections used at Y-12 are as follows:

For bioassay sample results less than the detection limit for the analytical technique, no action is 

required. Note, that corrections for samples from the early days when MDAs were significantly larger 

are not required.

To adjust results for background, the 234U-to238U ratios are reviewed for each sample to determine if 

the sample is consistent with EU or DU. Conservative ratios are applied in this determination: All 

samples with 234U-to238U ratios greater than 1:1 are considered to be consistent with EU, and all 

samples with 234U-to238U ratios less than 1:1 are considered to be consistent with DU.

The background contribution is subtracted from EU samples as follows: 

a. Assume that the ratio of 234U to 238U is 1:1 for dietary uranium. 

b. Determine if the 238U result exceeds its corresponding critical level value. 

c. If the 238U result is greater than the critical level, then subtract two times that 238U result from 

the total uranium result. 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 28 of 44

d. If the 238U result is less than its corresponding critical level value, do not subtract a dietary 

component.

For DU samples (e.g., those with a 234U-to238U ratio less than 1:1), the following steps are used to 

determine the dietary correction for DU samples:

a. The following equations are solved to determine an equation for the occupational 238U content 

and the occupational 234U content in a DU sample:

1 238

234

=

diet

diet

U

U (3-2)

2.0 238

234

=

occ

occ

U

U

(3-3)

actual diet UUU occ

238 238 238 += (3-4)

actual diet UUU occ

234 234 234 += (3-5)

b. Using Equation 3-5, and substituting the 238Udiet for the 234Udiet based on Equation 3-2 and 

substituting 0.2 (238Uocc) based on Equation 3-3 for the 234Uocc, the following relationship 

results: 

( ) act UU diet Uocc

234 238 238 += 2.0 (3-6)

Because the 238Udiet can be replaced with 238Uactual – 238Uocc derived from Equation 3-4, 

Equation 3-5 can be solved to determine the occupational quantity of 238U. The resultant 

equation for determining the occupational quantity of 238U is:

occ

act act U UU 238

234 238

8.0 = −

− (3-7)

c. Once the occupational 238U quantity is known, the occupational quantity of 234U can be derived 

using Equation 3-3:

( ) Uocc Uocc

234 238 = 2.0 (3-8)

d. Ensure that both the 234U and 238U results exceed their corresponding critical level values. If 

both of these results are not greater than their corresponding critical level values, then a 

background determination cannot be made. In this case, it is assumed that all of the activity 

observed in the total uranium sample result will contribute to occupational dose.

e. Use either Equation 3-7 or 3-8 to determine the occupational 238U component or the 

occupational 234U component, respectively. For ease of reference, the 238U component was 

chosen for the remaining steps.

f. Subtract the occupational component of the sample from the corresponding observed 

component to determine the dietary contributor:
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 29 of 44

act occ UUU diet

238 238 238 =− (3-9)

g. Once the dietary component has been calculated, subtract twice this dietary value from the 

total uranium result to remove the background:

( ) tot − diet = UUU totalocc

238 2 (3-10)

5.3 IN VIVO MINIMUM DETECTABLE ACTIVITIES, COUNTING METHODS, AND 

REPORTING PRACTICES

5.3.1

Whole-body counting was not routinely practiced at Y-12. The primary in vivo detection method was 

chest counting, as described below.

Whole-Body Counting

5.3.2

The Y-12 in vivo chest-counting facility was developed in the late 1950s and was put into routine use 

in 1961. The original facility is described in Cofield (1959) and Cofield (1960). A 9- by 4-in. sodium 

iodide (thallium doped) [NaI(Tl)] crystal was placed over the chest of a subject reclining in a cot in a 

shielded room. Subjects showered, shampooed, and changed into a clean garment before counting. 

They were then surveyed for any remaining surface contamination before entering the counting room. 

The normal counting time was 20 min.

Chest Counting

Incremental improvements in the system were made over time. A second 9- by 4-in. NaI(Tl) crystal 

was placed under the subject’s back in 1963 or 1964. It was found that this position provided greater 

counting efficiency than the position over the chest. A second advantage was that skin contamination 

seldom occurred on the back. The ratio of the count rates from the two crystals could be used to 

identify surface contamination missed by the survey meter. In 1965, two 5-in. NaI(Tl) crystals were 

added, positioned at the sides under the arms. In 1985, initial development of a high-purity 

germanium (HPGe) counting system was begun. The new system began routine operation in a new 

counting room in June 1992. 

5.3.2.1 Chest Counting for Uranium

With the NaI scintillation system used for much of the plant’s history, the amount of uranium deposited 

in the lungs was inferred from the worker’s spectrum by use of a prediction equation. This included 

the subject’s weight and constants derived from the spectra of control subjects. Compton scatter from 

natural 40K and fallout 137Cs in the body was a particularly important contributor to background in the 

235U region of interest around 186 keV and the 238U region of interest around 93 keV. The 137Cs 

contribution to the regions of interest changed over time as fallout levels varied, requiring frequent 

updates to the control subject spectrum library. 

Measured lung burdens were originally expressed in micrograms of 235U for EU and milligrams of 238U 

for NU or DU. The original limits of detection for 235U and 238U were reported as 130 μg and 13.5 mg, 

respectively (Cofield 1959). Improvements in hardware and data reduction procedures led to lower 

detection limits, as indicated in Tables 5-11 and 5-12 below. Limits are given in terms of both mass 

units and of activity (nanocurie and dpm), with the published value in regular type and the converted 

value in italics. When known, the reporting convention used to define the detection limit is indicated in 

the tables. The alternative definitions listed by King and Barclay (1983) illustrate the importance of 

this factor. 
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Due to the uncertainty about actual methods for determining the lung counting detection limits, for 

dose reconstruction purposes and based on review of the Y-12 data, the 235U and 238U lung count 

detection thresholds are assumed to be 130 μg and 13.5 mg, respectively, through 1990. Beginning 

in 1991, the critical level or MDA is supplied with the individual results. 

Table 5-11. Reported lung counting detection limits for 235U. 

Year Detectors

U-235 

(μg)

U-235 

(nCi)

U-235 

(dpm)

Reporting 

convention Reference

1959 One 9-in. NaI 130 0.28 620 Cofield 1959

1963 Two 9-in. NaI 96 0.21 470 Scott and West 1967

1965 Two 9-in., 2 5-in. NaI 72 0.16 360 Scott and West 1967

1975 Two 9-in., 2 5-in. NaI 70 0.15 330 Scott and West 1975

1983 Two 9-in., 2 5-in. NaI 62 0.13 290 1.96 σBKG King and Barclay 1983

1983 Two 9-in., 2 5-in. NaI 68.5 0.15 330 LC

a King and Barclay 1983

1983 Two 9-in., 2 5-in. NaI 137 0.30 670 LD

a King and Barclay 1983

1990 Two 9-in., 2 5-in. NaI 70 0.15 330 b Barber and Forest 1995

6/1992 HPGe 46 0.10 220 c BWXT Y-12 2003b

a. As defined by Currie (1968) with α = β = 0.05, paired observations.

b. Type I error = 5%.

c. Individual MDA or critical level supplied with analytical results.

Table 5-12. Reported lung counting detection limits for 238U. 

Year Detectors

U-238

(mg)

U-239 

(nCi)

U-238

(dpm)

Reporting 

convention Reference

1959 One 9-in. NaI 13.5 4.5 10,000 Cofield 1959

1965 Two 9-in., 2 5-in. NaI 6.5 2.2 4,900 Scott and West 1967

1983 Two 9-in., 2 5-in. NaI 5 1.7 3,800 1.96 σBKG King and Barclay 1983

1983 Two 9-in., 2 5-in. NaI 7.2 2.4 5,300 LC a King and Barclay 1983

1983 Two 9-in., 2 5-in. NaI 14.4 4.8 11,000 LD a King and Barclay 1983

1991 HPGe 4.5 1.5 3,300 b BWXT Y-12 2003b

a. As defined by Currie (1968) with α = β = 0.05, paired observations.

b. Individual MDA or critical level supplied with analytical results.

No description of the conversion count rate to activity to mass has been found. The analyte reported 

was based on the area in which the employee worked. Individuals working in NU or DU areas had 

results reported as 238U, and workers in enriched areas had results reported as 235U. 

Claimant-favorable assumptions should be based on conversions of 93% enrichment for 235U and 

natural isotopic abundances for 238U.

5.3.2.2 Chest Counting for Other Actinides

Thorium

The in vivo lung count was the only monitoring technique for monitoring thorium exposure in the body 

during the plant’s first decades. Thorium lung activity was inferred from 228Ac and/or 212Pb lung 

activity. Thorium lung counting was conducted from 1958 to 1984 with routine lung counts, scheduled 

at approximately six-month intervals, starting in 1961. 

The interpretation of thorium spectra is complicated by the tendency of decay chain members to 

become separated during operations, particularly those involving heat. The boiling point of radium is 

below the melting point of thorium. The 228Ac gamma radiation in an in vivo count could be an 

indication of a 228Ra intake or an intake of the entire 232Th chain. The 239-keV gamma line from 212Pb, 

a progeny of 228Th, is a more reliable indicator of 232Th, as the thorium isotopes would remain together 

during processing. However, if the 228Ra is lost, 212Pb decreases with the 1.9-yr half-life of its parent, 

228Th, moderated by the ingrowth of new 228Ra with its 5.7 yr half-life. This complicated disequilibrium 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 31 of 44

pattern operates both before and after chain members are taken into the body. Some knowledge of 

the likely state of equilibrium is therefore necessary to translate observed activities into dose.

The maximum permissible lung burden (MPLB) of 232Th (corresponding to 15 rem/yr to the lung) 

varied markedly with the ratio of 228Th to 232Th. For full equilibrium (ratio = 1) the MPLB was 

calculated to be 2.9 nCi of 232Th. For a ratio of 0.1, the lung burden was 12 nCi (West 1965, p. 22). 

For a ratio of 0.8, an MPLB of 3.2 nCi is indicated. For this ratio, with a 228Ra-to232Th ratio of 0.6, an 

MDA of 0.2 lung burdens, or 0.6 nCi, is indicated (West 1965, p. 26). In mass units, this is 5.5 mg

It was reported in 1965 that thorium at Y-12 was processed less than 1 yr after purification by the 

supplier and as a consequence had only about 10% as much 228Ra as 224Ra (West 1965, p. 18). This 

means that the maximum dose conversion factor per milligram of 232Th would be less than that for 

232Th in full equilibrium with its progeny. However, unless specific information is available in the 

claimants’ data, the dose reconstructor must make claimant-favorable assumptions. The thorium 

sensitivity varied due to the dependence of this technique on the ratios of 232Th and 228Ra to the 

progeny being measured. Data interpretation was based on a careful evaluation of the work histories. 

In the absence of sufficient data to determine the 228Th-to232Th ratios and state of equilibrium, the 

claimant-favorable assumption is to assume full equilibrium. 

. 

Neptunium

At the time the in vivo system was put into routine service in 1961, the reported detection limit for 

237Np without progeny radiation was 2.7 nCi (Cofield 1961). For 237Np in full equilibrium with 233Pa, the 

detection limit was reported as 0.255 nCi (Cofield 1961).

5.3.2.3 Chest Counting for Other Radionuclides

At the time the in vivo system was put into routine service in 1961, the reported detection limit for 60Co 

was 0.66 nCi. For 95Zr in transient equilibrium with 95Nb, the reported detection limit was 1 nCi. 

Although bremsstrahlung counting could have been done for 

5.4 UNMONITORED DOSE (NOT COMPLETE FOR ALL OPERATIONS AND/OR 

RADIONUCLIDES)

99Tc, no information is available about 

the sensitivity of the technique.

There are two eras of operations at Y-12 for which sufficient monitoring information might not be 

available in the worker records. From 1943 to 1947, limited or no monitoring was performed. No data 

have been found for this period. The primary site activities for this period were uranium enrichment by 

the Calutron process. From 1948 to 1950, fluorometric analyses of urine and blood were conducted 

as part of general medical surveillance to prevent kidney damage from exposure to soluble uranium 

compounds; these data cannot be found at this time. Coworker data will be applied to this time 

period. 
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 32 of 44

REFERENCES

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Oak Ridge, Tennessee, March 5.

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Health Physics, volume 2, pp. 269–287.

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Ridge National Laboratory, Oak Ridge, Tennessee, August 5.

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Nuclear Company, Oak Ridge, Tennessee, 1958.

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Permissible Dose for Internal Radiation (1959), Publication 2, Pergamon Press, Oxford, 

England.

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by Workers, Publication 30, Part 1, Pergamon Press, Oxford, England.

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International Commission on Radiological Protection, Publication 60, Pergamon Press, Oxford, 

England.

ICRP (International Commission on Radiological Protection), 1994, Human Respiratory Tract Model 

for Radiological Protection, Publication 66, Pergamon Press, Oxford, England. 

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

Publication 71, Pergamon Press, Oxford, England. 

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Exposure of Workers, Publication 78, Pergamon Press, Oxford, England.

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Electrodeposition Procedure of Enriched Uranium Urinalysis,” memorandum to P. R. 

Ammann), October 25, , Oak Ridge, Tennessee.

King, A., and R. K. Barclay, 1983, Calibration and Use of a Lung Monitoring Facility Using Sodium 

Iodide Detectors, Y-DD-317, Union Carbide Corporation, Oak Ridge, Tennessee, May.

McLendon, J. D., 1960, “Personnel Radiation Exposure Data,” memorandum to G. W. Flack (August 

25, , Oak Ridge, Tennessee.

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Security Complex – Site Description, ORAUT-TKBS-0014-02, Revision, Oak Ridge, 

Tennessee, November. 

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Environmental Management Project (FEMP) – Occupational Internal Dose, 

ORAUT-TKBS-0017-05, Revision, Oak Ridge, Tennessee, November. .
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Patterson, G. R., 1958, “Evaluation and Control of Internal Exposure From Enriched Uranium at 

Y-12,” in Symposium on Occupational Health Experience and Practices in the Uranium 

Industry, HASL-58, , October 15 to 17, U.S. Atomic Energy Commission, Health and Safety 

Laboratory, New York, New York.

Polednak, A. P., and Frome, E. L., 1981, “Mortality Among Men Employed Between 1943 and 1947 at 

a Uranium Processing Plant,” Journal of Occupational Medicine, volume 23, no. 3,

pp. 169–178.

Potter C. A., 2002, “Intake Retention Functions Developed from Models Used in the Determination of 

Dose Coefficients Developed for ICRP Publication 68—Particulate Inhalation,” Health Physics

volume 83, pp. 594–789.

Rucker T. L., J. D. Slack, K. N. Fleming, S. W. Stevens, R. M. Moody, C. M. Johnson, and S. W. 

Green, 2001, "Uranium Lung Solubility Class Selection at Bechtel Jacobs Company 

LLC-Operated Facilities," presented at Radiochemical Measurements Conference, Honolulu, 

Hawaii, November 4 to 8. 

Scott, L. M., 1964, “Excretion of Uranium from Mixed Exposures Under Industrial Conditions,” Health 

Physics, volume 10, pp. 479–481.

Scott, L. M., and C. M. West, 1967, Health Physics Application of In Vivo Gamma Spectrometry in a 

Uranium Processing Plant, Y-KB-77, Union Carbide Corporation, Oak Ridge, Tennessee.

Scott, L. M., and C. M. West, 1975, Uranium In Vivo Spectrometry, Y-DD-173, Union Carbide Nuclear 

Company, Oak Ridge, Tennessee, April, 1975.

Snapp, L., 2003, “Case Studies of Long Retention Times”, memorandum to H. Prichard, September 

9,.

Souleyrette, M. L., 2003, Summary of Historical Monitoring Techniques Provided to NIOSH for 

EEOICPA Data Requests, BWXT Y12, Oak Ridge, Tennessee, February 18, 2003.

Steckel, L. M., and C. M. West, 1966, Characterization of Y-12 Uranium Process Materials Correlated 

with In Vivo Experience, Y-1544-A, Union Carbide Corporation, Oak Ridge, Tennessee.

Sterner, J. H., and E. C. Riley, 1946, “Number of Overexposures to Hazards noted in Chart B,” 

document fragment labeled “From Dr. J. Sterner, Eastman Kodak”, located in ORAU 

Document Vault, Eastman Kodak Corporation, Oak Ridge, Tennessee.

UCC (Union Carbide Corporation), 1949, Health Physics Progress Report: January 1 - 31, 1949, Oak 

Ridge, Tennessee, February 14.

UCC (Union Carbide Corporation), 1950, Health Physics Progress Report: November 1 - December 

31, 1950, Oak Ridge, Tennessee, June 14.

UCC (Union Carbide Corporation), 1952, Health Physics Progress Report: July 1 to December 31 

1951, Oak Ridge, Tennessee, March 15.

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Tennessee.
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UCC (Union Carbide Corporation), 1963a, Y-12 Radiation Safety Manual, July 23, 1963, Y-KB-29, 

Nuclear Division, Oak Ridge, Tennessee.

UCC (Union Carbide Corporation), 1963b, Y-KB-33, Nuclear Division, Oak Ridge, Tennessee.

UCC (Union Carbide Corporation), 1965, Y-12 Radiation Safety Manual March 24, 1965, Nuclear 

Division, Oak Ridge, Tennessee.

UCC (Union Carbide Corporation), 1966, Analysis of Urine for Enriched Uranium by Electrodeposition, 

Nuclear Division, Oak Ridge, Tennessee.

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Procedures, Oak Ridge, Tennessee, December 11, 1959.

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Union Carbide Corporation, Oak Ridge, Tennessee, January 19.

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Y-12 Plant, Y-KB-19, Union Carbide Nuclear Company, Oak Ridge, Tennessee, September.

West, C. M., and L. M. Scott, 1966, “A Comparison of Uranium Cases Showing Long Chest Burden 

Retentions,” Health Physics, volume 12, pp. 1545-1555.

West, C. M., and L. M. Scott, 1969, “Uranium Cases Showing Long Chest Burden Retention - An 

Updating,” Health Physics, volume 17, pp. 781–791.

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Chronology of Some Noteworthy Events and Contributions During Y-12’s First Fifty Years by a 

Group of Participants in that History”
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GLOSSARY

activity median aerodynamic diameter (AMAD)

The diameter of a sphere having a density of 1 gram per cubic centimeter and the same 

terminal settling velocity in air as that of the aerosol particle whose activity is the median for 

the entire aerosol.

acute exposure

Exposure of short duration.

background (radiation)

Radiation from:

• Naturally occurring radioactive materials that have not been technologically enhanced

• Cosmic sources

• Global fallout as it exists in the environment (from the testing of nuclear explosive devices)

• Radon and its progeny in concentrations or levels existing in buildings or the environment 

that have not been elevated as a result of current or previous activities

• Consumer products that contain nominal amounts of radioactive material or produce 

nominal amounts of radiation.

bioassay

The determination of kinds, quantities, or concentrations and, in some cases, locations of 

radioactive material in the human body, whether by direct measurement or by analysis and 

evaluation of radioactive materials excreted or removed from the human body. Also called 

radiobioassay. 

chronic exposure

Exposure of long duration, delivered by fractionation or protraction.

class D, W, or Y material

ICRP 30 classifies inhaled radioactive materials as type D, W, or Y (days, weeks, or years) 

depending on their retention times in the pulmonary region. Class D materials have a 

pulmonary half-life of less than 10 d; W materials have a half-life from 10 to 100 d; and Y 

materials have a half-life greater than 100 d. The times actually used for these classes of 

materials in ICRP 30 in the calculation of annual limits on intake are 0.5 d, 50 d and 500 d for 

class D, W, and Y materials, respectively. Note: A class Y uranium has been defined as 

having an effective pulmonary half-life of 100 d and the systemic distribution, retention, and 

excretion parameters of class W uranium. 

class (type) Q system

A class Q dosimetry system uses an 8-μm AMAD as a basis. The class Q dosimetry system is 

described as 10% class Y and 90% modified class W (the modification being an increase of 

the class W 50-d compartment to 120 d).

committed effective dose equivalent (CEDE)

The sum of the committed dose equivalents to various tissues or organs in the body for 50 yr 

after intake. CEDE is expressed in units of rem (or sievert).
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depleted uranium (DU)

A byproduct of the enrichment process, depleted uranium has a reduced abundance of 235U 

relative to the 0.72% abundance in natural uranium. 

dose

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

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

enriched uranium (EU)

Uranium with an increased abundance of 235U. Enriched uranium varies from greater than the 

natural 235U abundance of 0.72% up to almost 100%. Enriched uranium can also contain 

increased abundances of 234U, 236U, 232U, 233U and others. 

exposure

(1) The general condition of being subjected to radiation, such as by exposure to radiation 

from external sources or to radiation sources inside the body. (2) In this document, exposure 

does not refer to the radiological physics concept of charge liberated per unit mass of air. The 

product of exposure time to a radioactive aerosol and the average concentration during 

exposure, divided by the value of the derived air concentration for the radioactive material in 

question (expressed in derived air concentration-hours). 

intake

The amount of radionuclide (mass or activity) taken into the body by inhalation, absorption 

through the skin, injection, ingestion, or through wounds.

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 measurements of radioactive material in the human body using instrumentation that 

detects radiation emitted from the radioactive material in the body.

maximum permissible lung burden (MPLB)

During the 1970s the occupational limit for plutonium was expressed in terms of a quantity of 

plutonium that could be present in the chest at any given time, equal to 16 μCi (0.25 μg) 239Pu.

minimum detectable amount or activity (MDA)

The smallest amount (mass) or activity (per mass or per sample) of an analyte in a sample 

that will be detected with a probability ß of nondetection (type II error) while accepting a 

probability of erroneously deciding that a positive (nonzero) 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

The measurement of radiation levels, airborne radioactivity concentrations, radioactive 

contamination levels, quantities of radioactive material, or individual doses.
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natural uranium

The mixture of uranium as it is found in nature (i.e., 99.27% 238U, 0.72% 235U, and 0.0054% 

234U by weight). The alpha activity comes mostly, and in about equal amounts, from 238U and 

234U. The specific activity of this mixture is 2.6 x 107 Bq/kg (0.7 pCi/g).

nephrotoxicity

Poisoning of the kidney.

quality factor

A modifying factor used to calculate the dose equivalent from the absorbed dose.

recycled uranium (RU)

Uranium recovered from spent fuel from fission reactors.

routine bioassay monitoring

Any bioassay measurement made on a predetermined, periodic schedule, to establish a 

worker’s internal exposure status relative to previous periods of time.

sensitivity (detection sensitivity)

The minimum amount of contaminant (or dose) than can be repeatedly measured (or 

calculated) by a particular analysis. See minimum detectable amount. 

type F, M, or S material

ICRP 66 classifies inhaled radioactive materials as F, M, or S (fast, medium, or slow) 

depending on their retention time in the pulmonary region. These designations are similar to 

the clearance classes D, W, and Y, but refer strictly to the rate of absorption from the lungs to 

the blood.
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ATTACHMENT 5A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

TABLE OF CONTENTS

Section

5A.1 Conventions Used in Internal Dose Reports..............................................................................

Page

40

5A.1.1 Y-12 Urinalysis Data through September 1989................................................................. 40

5A.1.2 Y-12 Urinalysis Data since October 1989 .........................................................................40

5A.1.3 In Vivo Data......................................................................................................................41

5A.2 Uranium Solubility in the Lung................................................................................................... 43

5A.3 In Vitro Detection Limits ............................................................................................................ 43

5A.4 In Vivo Detection Limits............................................................................................................. 43

LIST OF TABLES

Table

5A-1 Uranium urinalysis detection limits .........................................................................................

Page

43

5A-2 Other in vitro detection limits .................................................................................................. 43

5A-3 Reported in vivo detection limits for other radionuclides......................................................... 44
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5A.1 CONVENTIONS USED IN INTERNAL DOSE REPORTS 

The following is adapted from Souleyrette (2003). 

5A.1.1 Y-12 Urinalysis Data through September 1989

Dept – Department number of employee at the time of monitoring.

Badge/Old SSN – Employee identification number.

Void Date – Date sample was voided.

1 = EU urinalysis by gross alpha

Program Code 

2 = NU or DU by fluorophotometry (used for EU before 1950)

6 = control sample

9 = (may have been used for a limited time to indicate experimental analyses methods).

Volume – volume of sample in ml.

Time – Time between last void and sample void. Used to normalize urine result to 24-hr void based 

on individual employee daily excretion rate. (A zero time in conjunction with a large void volume and 

program code 9 probably indicates an experimental measurement technique.)

Uranium Result (dpm) – The normalized urine result in disintegrations per minute per day.

Background Value 

5A.1.2

– Background was stored for later results. The column Uranium Result (dpm) is 

background corrected already.

Y-12 Urinalysis Data since October 1989

Dept - Department number of employee at the time of monitoring.

Badge - Employee identification number.

Sample Number – Laboratory sample identifier.

Sample Date - Date sample was voided.

Sample Type

A = Alpha spec 

 

F = fecal

U = urine 

S = Start or stop date 

(Note: For urinalysis, A is expected because all urinalyses were done this way since October 1989. 

The other entries are for database flexibility.)
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Reason

1S = acute exposure 

 – Reason for scheduling a sample: 

2S = follow-up sample to an acute exposure 

3 = routine (chronic) exposure

2R = follow-up to routine (chronic) exposure

4 = background sample 

5 = baseline sample 

6 = random sample 

8 = start date of exposure * 

9 = stop date of exposure *

0 = acute exposure

* (These should appear only in separate records introducing a series related to a chronic intake 

study).

Retention Class – Assigned retention class: ICRP-30 classes D, W, and Y. (Class Q indicates 90% 

class Super–W (120-d lung retention) and 10% class Y, 8 μm AMAD.)

Volume (ml) or Weight (g) – sample volume (urine) or weight (fecal).

Total Uranium Result (dpm) - This is the normalized total uranium urine result in disintegrations per 

minute per day.

Type Assimilation 

H = inhalation 

 

I = ingestion

J = injection

Incident Date – Date of documented radiological exposure incident, if any.

Incident Time

5A.1.3

 – Time of documented radiological exposure incident, if any.

In Vivo Data

Dept – Department number of employee at the time of monitoring.

Run Date – Date subject counted.

Run No. – Laboratory identifier of counting session.

Weight (lbs.) – Subject weight in pounds.

Chest Thickness (in.) – Subject chest wall thickness in inches (could be in tenths of inches for some 

years).

Surface Contamination Code

• If the first position is a 1, there was no surface contamination found, and the remaining five 

digits should be zeros. 

 – A six-digit code indicating whether or not surface contamination was 

detected. 
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• If the first digit is a 2 or a 3, there was surface contamination found and the remaining five 

digits give the survey meter reading of the contamination.

Type Analysis

From October 1962 to 1971, the codes are:

 – Code indicating the activity regions analyzed. There are two distinct sets of codes 

depending on the date of the count.

0 = Background run

1 = Uranium 

2 = New hires/control counts 

3 = Thorium

4 = Neptunium

5 = Miscellaneous

6 = Second thorium count.

Since 1972, the codes are:

1= 235U 

3 = Thorium

4 = 238U and 235U 

6 = Thorium and 235U (first thorium count)

7 = 235U and Thorium (second thorium count).

Material Type

For counts after 1971, where Type Analysis is 0:

 – This code provides additional details for cases where Type Analysis = 0, 1, 2, or 3.

00 = Background run

01 = Sensitivity Check.

For Type Analysis of 1 (any year) or Type Analysis of 2 (after 1971):

01 = Unknown

02 = DU

03 = DU

07 = Normal uranium

13 = EU

19 = EU.

For Type Analysis of 3 (any year):

00 = Thorium (first exam)

02 = Thorium (second or later exam).

U-235 (μg) – 235U activity detected in chest cavity. If the count is invalid, an ‘&’ appears as the first 

character.

Thorium (mg) – 232Th activity calculated from the daughter isotopes detected in chest cavity.

U-238 (mg) – 234Th activity detected in chest cavity.
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Tc-99 (μCi) - 99Tc activity detected in chest cavity.

Current Dept

5A.2 URANIUM SOLUBILITY IN THE LUNG

 – Additional instance of employee department number used historically for reporting 

purposes.

For a workplace as varied as Y-12, it is clear that no single solubility type or particle size distribution 

applies to all workers. Furthermore, accurate assignment of the uranium lung clearance type to a 

given bioassay result is considered virtually impossible because of uncertainties about chemical form 

and limitations of the personnel-tracking system (Barber and Forest 1995, p. 669). Exposure to type 

M material from 1948 to June 1998 appears to be the most likely absorption type. After June 1998, 

exposure to absorption type S material is more likely. However, the absorption type can be based on 

the monitoring data or claimant-favorable assumptions. 

5A.3 IN VITRO DETECTION LIMITS

Tables 5A-1 and 5A-2 summarize information developed in Section 5.2. The tabulated values for 

urinalysis results represent laboratory detection limits LD and do not include uncertainties introduced 

by sample collection or conversion from submitted volumes to daily void volumes. As noted in 

Section 5.2, LD values for some historical techniques remain to be identified, and will be reported in 

subsequent revisions as available.

Table 5A-1. Uranium urinalysis detection limits. 

Method Period

Detection 

limit (mass)

Detection 

limit (activity) Convention 

Fluorometry 1950–9/1989 7 μg/d 11 dpm/d

Gross alpha 1950–1964 47 dpm/d LD

Gross alpha 1965–9/1989 26 dpm/d LD

Alpha spectrometry 10/1989–present 0.15 dpm/d LD

Table 5A-2. Other in vitro detection limits. 

Analyte/solubility type Method Period

Detection limit 

(activity)

Tritium (HTO) Liquid scintillation 10/1988–present 2,000 dpm/da

Isotopic plutonium/M, S Alpha spectrometry 10/1989–present 0.025 dpm/sample

Am-241/M Alpha spectrometry 10/1989–present 0.050 dpm/sample

Th-228/M, S Alpha spectrometry 10/1989–present 0.150 dpm/sample

Th-232/M, S Alpha spectrometry 10/1989–present 0.070 dpm/sample

Np-237/M Alpha spectrometry 10/1989–present 0.100 dpm/sample

a. Estimate.

5A.4 IN VIVO DETECTION LIMITS

From the review of Y-12 documents and conversations with present and former Y-12 staff, the vast 

majority of site operations from 1943 to the present involved uranium in a variety of chemical forms 

and degree of enrichment. Therefore, it is concluded that the primary internal radiation exposure to 

Y-12 workers was from uranium. However, the internal dosimetry program has included limited 

monitoring for 137Cs, 99Tc, thorium, plutonium, 228Ac, and tritium, among other radionuclides. There 

are difficulties in interpreting these measurement data that cannot be resolved satisfactorily at this 

time. These issues should not have a large impact on the worker’s internal dose because exposure to 

uranium is the source of the greatest part of the internal dose.
Effective Date: 05/10/2005 Revision No. 01 Document No. ORAUT-TKBS-0014-5 Page 44 of 44

Table 5A-3 summarizes information developed in Section 5.3. Detection limits for some historical 

techniques remain to be identified, and will be reported in subsequent revisions as available. 

Table 5A-3. Reported in vivo detection limits for other radionuclides. 

Radionuclide

/solubility type

Reported limit 

(nCi)

Reported limit 

(dpm)

Citation 

year

Th-232/M, S 0.6 1,300 1965

Np-237/M 2.7 5,900 1961

Np-237/M–Pa-233/M, S 0.255 560 1961

Co-60/M, S 0.66 1,500 1961

Zr-95/M, S–Nb-95/M, S 1.0 2,200 1961

Although internal monitoring has existed from the earliest days, the data from these measurements 

are not available for years before 1950. Guidance for the period from 1948 to 1950 is provided, but it 

is not feasible to provide any guidance for exposures before 1948. This will be provided in revisions 

to this document.

For dose reconstruction purposes and based on review of the Y-12 data, the 235U and 238U lung count 

detection thresholds are assumed to be 130 μg and 13.5 mg, respectively, through 1990. Beginning 

in 1991, the critical level or MDA is supplied with the individual results.