Draft

A REVIEW OF ORAUT-OTIB-0080, REV. 00: 

 INTERNAL COWORKER DOSIMETRY DATA FOR AREA IV OF THE 

SANTA SUSANA FIELD LABORATORY AND THE DeSOTO AVENUE 

FACILITY 

Contract No. 211-2014-58081 

SCA-TR-OTIB2014-0080, Revision 0 

Prepared by 

Bob Barton 

Joyce Lipsztein 

Milton Gorden 

S. Cohen & Associates 

1608 Spring Hill Road, Suite 400 

Vienna, VA 22182 

November 2014 

Disclaimer 

This document is made available in accordance with the unanimous desire of the Advisory Board on 

Radiation and Worker Health (ABRWH) to maintain all possible openness in its deliberations. However, 

the ABRWH and its contractor, SC&A, caution the reader that at the time of its release, this report is predecisional and has not been reviewed by the Board for factual accuracy or applicability within the 

requirements of 42 CFR 82. This implies that once reviewed by the ABRWH, the Board’s position may 

differ from the report’s conclusions. Thus, the reader should be cautioned that this report is for 

information only and that premature interpretations regarding its conclusions are unwarranted.
Effective Date:

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However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

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S. Cohen & Associates:

Technical Support for the Advisory Board on

Radiation & Worker Health Review of

NIOSH Dose Reconstruction Program

Document No.: 

SCA-TR-OTIB2014-0080

Effective Date: 

Draft – November 24, 2014

Revision No.:

0 (Draft)

A REVIEW OF ORAUT-OTIB-0080, REV. 00: 

INTERNAL COWORKER DOSIMETRY DATA 

FOR AREA IV OF THE SANTA SUSANA FIELD 

LABORATORY AND THE DeSOTO AVENUE 

FACILITY

Page 2 of 69

Task Manager:

_____________________________ Date: ___________

John Stiver, MS, CHP

Supersedes:

N/A

Project Manager:

_____________________________ Date: ___________

John Stiver, MS, CHP

Reviewers:

 

 John Stiver

 

Record of Revisions

Revision 

Number

Effective 

Date Description of Revision

0 (Draft) 11/24/2014 Initial issue
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However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

TABLE OF CONTENTS 

Abbreviations and Acronyms ..........................................................................................................6

Executive Summary .........................................................................................................................8

1.0 Characterization of Available Coworker Dataset ..............................................................13

2.0 Overview of Monitored Job Types Contained in ORAUT-OTIB-0080 Database ............21

3.0 Review of Plutonium Data and Associated Coworker Intakes ..........................................24

3.1 Review of Calculated OPOS Excretion Results ....................................................24

3.2 Replication of Calculated Plutonium OPOS Values ..............................................28

3.3 Discussion of Historical Plutonium Processes .......................................................29

3.4 Occupations Monitored for Plutonium Intake .......................................................33

4.0 Review of Fission Product Data and Associated Coworker Intakes..................................35

4.1 Review of Calculated OPOS Results .....................................................................35

4.2 Replication of Calculated OPOS Results ...............................................................37

4.3 Occupations Monitored for Fission Product Intake ...............................................39

5.0 Review of Uranium Data and Associated Coworker Intakes ............................................41

5.1 Review and Reproduction of Calculated OPOS Results .......................................41

5.2 Review of Uranium Monitoring Practices and Incorporation into Internal 

Model .....................................................................................................................44

5.3 Exposures to Uranium Aluminide .........................................................................50

5.4 Occupations Monitored for Uranium Intake ..........................................................52

6.0 Evaluation of Need for Modeling Other Constituents .......................................................55

References ......................................................................................................................................60

Attachment 1: Breakdown of Internal Monitoring Entries Contained in the 

“Nuclide/Analysis” Column of the Coworker Database ...................................................63

Attachment 2: List of Uranium Fire Incidents ..............................................................................64

Attachment 3: Occupational Categorization .................................................................................69
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LIST OF TABLES 

Table 1. Description and Rationale for the Removal or Extraction of Entries from the 

Database for Coworker Analysis ................................................................................. 13

Table 2. Overview of Bioassay Sample Units Available for Coworker Analysis ....................... 14

Table 3. Percentage of Entries for Each Contaminant with Known and 

Unknown Sample Volumes ......................................................................................... 14

Table 4. Characterization of Paired Results in which the “Reported” Value was 

Designated as “Less Than” .......................................................................................... 20

Table 5. Occupations Monitored at SSFLa .................................................................................. 23

Table 6. Plutonium Excretion Rates (dpm/1.4L) ......................................................................... 24

Table 7. Plutonium OPOS Analysis Period Reconstruction Based on Effective Bioassay 

Dates Shown in ORAUT-OTIB-0080 Table A-1 ........................................................ 25

Table 8. Comparison of SC&A-Calculated OPOS Results and ORAUT-OTIB-0080 

Table A-1 OPOS Results ............................................................................................. 28

Table 9. Operations at the Nuclear Materials Development Facility 

Operating Period Operation ......................................................................................... 29

Table 10. Bioassay Data for the NMDF ...................................................................................... 31

Table 11. Plutonium Bioassay Results from 1989–1999 ............................................................. 32

Table 12. Occupations Monitored for Plutonium Intakes at SSFLa............................................ 34

Table 13. Comparison of SC&A and NIOSH OPOS Totals by Year .......................................... 35

Table 14. Additional Sample Types Not Meeting the Criteria for Inclusion per 

ORAUT-OTIB-0080 .................................................................................................... 36

Table 15. Comparison of Expanded SC&A OPOS Totals and NIOSH OPOS 

Totals by Year .............................................................................................................. 37

Table 16. Comparison of SC&A-Calculated OPOS Values Using Expanded Fission 

Product Data to NIOSH OPOS Values contained in Table A-3 of ORAUTOTIB-0080 ................................................................................................................... 37

Table 17. Occupations Monitored for Mixed Fission Product Intakes at SSFLa ......................... 40

Table 18. Comparison of SC&A-Calculated OPOS Results to NIOSH-Calculated OPOS 

Results Contained in ORAUT-OTIB-0080 Table A-2 ................................................ 41

Table 19. Tabulation of Available 1984 and 1985 Data .............................................................. 43

Table 20. Description of Bioassay Sampling Protocol at SSFL .................................................. 45

Table 21. Distribution of Bioassay Results by the Day of the Week Sample is Collected ......... 45

Table 22. Distribution of Bioassay Results by “Sample Received” Date .................................... 46

Table 23. Predicted Urine Excretion Rates Corresponding to NIOSH’s Derived GM Intake 

Rates of Type F Uranium Compounds, Assuming a Continuous Exposure as 
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compared to a 5-day Per Week Exposure, with Monday Morning Sample 

Collections ................................................................................................................... 47

Table 24. Predicted Urine Excretion Rates Corresponding to NIOSH’s Derived GM Intake 

Rates of Type M Uranium Compounds, Assuming a Continuous Exposure as 

compared to a 5-day Per Week Exposure, with Monday Morning Sample 

Collections ................................................................................................................... 48

Table 25. Predicted Urine Excretion Rates Corresponding to NIOSH’s Derived Geometric 

Mean Intake Rates of Type S Uranium Compounds, Assuming a Continuous 

Exposure as compared to a 5-day Per Week Exposure, with Monday Morning 

Sample Collections....................................................................................................... 49

Table 26. Occupations Monitored for Uranium Intakes at SSFLa ............................................... 53

LIST OF FIGURES 

Figure 1. Percentage of Records Listing a Sample Volume for Uranium, 

Plutonium and Fission Products ................................................................................... 16

Figure 2. Example Logbook Contained in Claimant Hardcopy File Showing Units of 

“dpm/day” .................................................................................................................... 17

Figure 3. Corresponding Entry in Coworker Database: Entries Showing Units of “dpm” 

with a Blank Field for “Analyzed Volume” ................................................................. 17

Figure 4. Example Database Entry Showing “Reported” and “Calculated” Urinalysis 

Activity Results ............................................................................................................ 18

Figure 5. Example Bioassay Form Showing Both a “Calculated” Value and 

a “Reported” Value ...................................................................................................... 19

Figure 6. Less than “Reported” Results versus the Associated Sample Volume ........................ 19

Figure 7. Comparison of Available 1965 and 1966 Plutonium Urinalysis Values Including 

Interpolated Values at the 50th and 84th Percentile with Calculated OPOS Results 

from 1967–1968 at the 50th and 84th Percentile ........................................................... 27

Figure 8. Calculated NIOSH OPOS Values by Year ................................................................... 39

Figure 9. Comparison of the Number of Samples for Coworker Radionuclides 

during the 1980s ........................................................................................................... 43

Figure 10. Comparison of the Number of Monitored Workers for Coworker Radionuclides 

during the 1980s ........................................................................................................... 44

Figure 11. Predicted Uranium Bioassay Results Calculated using IMBA-derived Uranium 

Intake Rates (Line) compared with Bioassay Results (Dots), 50th Percentile, 

All Years, Type F ......................................................................................................... 47
Effective Date:

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ABBREVIATIONS AND ACRONYMS 

Advisory Board 

or ABRWH Advisory Board on Radiation and Worker Health 

AETR Advanced Epithermal Thorium Reactor 

AMAD Activity Mean Aerodynamic Diameter 

BZ breathing zone 

CEDE Committed Effective Dose Equivalent 

CEP Controls for Environmental Pollution, Inc. 

Ci curie 

cm2 square centimeter 

CMD count median diameter 

cpm count per minute 

d day 

DOE (U.S.) Department of Energy 

dpm disintegrations per minute 

ETEC Energy Technology Engineering Center 

EU enriched uranium 

FP fission product bioassay analysis method 

g/cm3 gram per cubic centimeter 

GB gross beta bioassay analysis method 

GIF Gamma Irradiation Facility

gm grams 

GM geometric mean 

GSD geometric standard deviation

HEPA high efficiency particulate air 

ICRP International Commission on Radiological Protection 

IMBA Integrated Modules for Bioassay Analysis 

kg kilogram 

L or l liter 

LETF Laser Experimental Test Facility 

LMEC Liquid Metal Engineering Center 

MDA minimum detectable amount 
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MFP mixed fission product bioassay analysis method 

MFPB or MFP(B) mixed fission product for beta counting bioassay analysis method 

MFPG mixed fission product for gamma counting bioassay analysis method 

MPC maximum permissible concentration 

μCi/cc microcurie per cubic centimeter 

μg micrograms 

μm micrometer 

mL milliliter 

MPM Maximum Possible Mean 

mrem millirem

nCi nanocuries 

NIOSH National Institute for Occupational Safety and Health 

NMDF Nuclear Materials Development Facility 

NOCTS NIOSH/OCAS Claims Tracking System 

NRC (U.S.) Nuclear Regulatory Commission 

OCAS Office of Compensation Analysis and Support 

OPOS One Person – One Sample

ORAUT Oak Ridge Associated Universities Team 

ORNL Oak Ridge National Laboratory 

PuO2 plutonium oxide 

RMDF Radioactive Materials Disposal Facility 

SC&A S. Cohen and Associates (SC&A, Inc.) 

SEC Special Exposure Cohort 

SRDB Site Research Database 

SRE Sodium Reactor Experiment 

SSFL Santa Susana Field Laboratory 

TBD Technical Basis Document 

TRUMP-S Transuranic Management by Propartitioning-Separation (TRUMP-S)

U-Alx uranium aluminide 

U3O8 triuranium octoxide 

UO2 uranium dioxide 

UR uranium bioassay analysis method 

UX unknown unit 
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EXECUTIVE SUMMARY 

The Santa Susana Field Laboratory (SSFL) is located approximately 30 miles northwest of 

downtown Los Angeles, California. SSFL consists of 2,850 acres in the Simi Hills of Ventura 

County. SSFL is divided into four administrative and operational portions based on ownership 

and operations. Department of Energy (DOE) operations are conducted in Rockwell 

International-owned and DOE-owned facilities in an area designated as Area IV. SSFL was 

initially established by North American Aviation in 1947 as a field test laboratory to static-fire 

rocket engines. In 1953, Area IV was established as a nuclear research and development facility 

with it being managed by Atomic International starting in 1955. Atomic International conducted 

activities in Area IV on development of civilian nuclear power, and the Liquid Metal 

Engineering Center (LMEC) that focused on research and testing of non-nuclear components to 

liquid metals. LMEC was renamed the Energy Technology Engineering Center (ETEC) in 1978. 

Through various mergers, the SSFL is now managed by Boeing. 

Between 1954 and 1980, several small nuclear reactors and critical test assemblies were built, 

tested, and operated in Area IV. These research facilities focused on development and operation 

of homogeneous water boiler-type reactors, sodium-cooled graphite-moderated reactors, and 

uranium-zirconium hydride reactors. Starting in 1956 and continuing through 1996, operations 

also supported the manufacture, management, and disassembly of nuclear reactor fuel, as well as 

the operation of nuclear waste management facilities. Fuel manufacturing included the assembly 

of fuel elements for the Sodium Reactor Experiment (SRE), a plutonium fuel manufacturing 

facility, and a uranium carbide fuel pilot plant. There was also a Fuel Storage Facility for 

storage of special nuclear materials used to make the fuels. Other research programs addressed 

reprocessing of used reactor fuel, operation of small particle accelerators, and use of 

radioisotopes. 

Three other facility locations are associated with SSFL operations: Downey, Canoga Park, and 

DeSoto. Downey and Canoga Park were operational prior to 1960 and are not considered in this 

report.1 At DeSoto, radiological operations occurred from 1959 into the mid-1990s. These 

operations included a Nuclear Regulatory Commission (NRC)-licensed research reactor, fuel 

fabrication for the Advanced Test Reactor, the Gamma Irradiation Facility that used sealed 

Cs-137 and Co-60 sources, and a mass spectroscopy laboratory. Some workers worked 

interchangeably between DeSoto and SSFL. 

1

 Special Exposure Cohort (SEC) Petitions 00093 and 00156 were designated in 2009 and 2010 

(respectively) for the Santa Susana operational period extending from January 1, 1955, through December 31, 1964, 

based on the inability to reconstruct internal doses. 

The technical information bulletin ORAUT-OTIB-0080 titled, Internal Coworker Dosimetry 

Data for Area IV of the Santa Susana Field Laboratory and the De Soto Avenue Facility

(ORAUT 2014a), presents an overview of the available internal monitoring data and derived 

coworker intake rates to be used at SSFL. SC&A was tasked by the Advisory Board on 

Radiation and Worker Health to review this document and associated coworker data for the 

feasibility of reconstructing unmonitored internal exposures. This report documents the results 

of that review. 

 
Effective Date:

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Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

Section 1 of this report provides an overview of the available dataset provided to SC&A, as well 

as the general characteristics of the applicable internal monitoring data, while Section 2 

summarizes how these data are distributed among occupations. Sections 3, 4, and 5 individually 

focus on the three contaminants of interest analyzed in ORAUT-OTIB-0080: “plutonium,” 

“fission products,” and “uranium,” respectively. Section 6 provides an evaluation of the need to 

model other radionuclides that would not have been monitored using analytical methods for 

uranium, plutonium, or mixed fission products. SC&A’s review focuses on the completeness 

and adequacy of the available data for the purposes of assigning coworker intake rates to 

unmonitored workers. In general, the “completeness” of a dataset refers to the extent to which 

the available data provide sufficiently representative and/or bounding temporal, work area, and 

job type coverage. “Adequacy” of the coworker dataset refers to the extent to which the 

numerical values used in intake calculations accurately reflect and/or bound the internal exposure 

potential experienced by the unmonitored worker population. 

Based on this review, SC&A has produced 15 findings as listed below: 

Finding 1: It is evident that the “reported” values with the “<” designation represent the 

volume-adjusted minimum detectable activity (MDA), and that the “calculated” values are much 

less than the censorship level. In these instances, NIOSH should utilize the “reported” value as a 

censored result, which is claimant favorable and in accordance with the Maximum Possible 

Mean (MPM) methodology in ORAUT-RPRT-0053 (ORAUT 2012).

Finding 2: SC&A was unable to reconcile apparent differences between the effective bioassay 

dates listed in Table A-1 of ORAUT-OTIB-0080 and the known start and end dates of 

“statistically usable” plutonium data. 

Finding 3: Combining multiple years for the purpose of calculating the 50th and 84th percentile 

One Person – One Sample (OPOS) results must be scientifically justified in the context of the 

potential for historical changes in operations and associated intakes for monitored workers. 

 

Finding 4: More definitive evidence is warranted to establish that exposure potential to 

plutonium was lower in 1965–1966 than it was in 1967–1968 to validate the back extrapolation 

of plutonium results to the earlier period. 

Finding 5: SC&A was not able to duplicate the plutonium OPOS values presented in Table A-1 

to within a reasonable level of precision. The source of the discrepancies should be identified 

and resolved prior to the calculation of coworker intakes. 

 

Finding 6: SC&A’s research into the historical use of plutonium at SSFL indicates that 

programs or campaigns in plutonium facilities did not extend for more than 1 or 2 years in most 

cases [at least in the case of the primary plutonium facility, Nuclear Materials Development 

Facility (NMDF)] and reinforces the need for Findings 3 and 4 to be addressed. Therefore, 

averaging plutonium data across multiple years for the purposes of calculating coworker intakes 

may not be appropriate. 
Effective Date:

November 24, 2014

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However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

Finding 7: Based on the number of monitored workers tabulated in Tables 13 and 14, it is likely 

that NIOSH did not adhere to the inclusion guidelines outlined in ORAUT-OTIB-0080. It 

appears that all samples labelled “FP” or “MFP” in the “Nuclide/Analysis” column were 

included regardless of the actual designated method of measurement. 

Finding 8: SC&A was not able to duplicate the “FP” OPOS values presented in Table A-3 of 

ORAUT-OTIB-0080 to within a reasonable level of precision for most years. The source of the 

discrepancies should be identified and resolved prior to the calculation of coworker intakes. 

 

Finding 9: The coworker model would benefit from a substantive discussion providing the 

rationale for combining all 27 years into a single intake regime. This might include a discussion 

on the variability in the calculated measurements observed by year, changes in the geometric 

standard deviation (GSD) indicating a larger spread of potential exposures, the actual effect on 

the calculated intakes in relation to dose, changes in site operations, etc. Such a discussion 

would reinforce the assumption that a single chronic intake regime is appropriate over a 

relatively long period of time. 

Finding 10: SC&A’s attempt to duplicate the uranium OPOS values presented in Table A-2 of 

ORAUT-OTIB-0080 was off by an order of magnitude in 4 of the 25 years analyzed by NIOSH 

(1984 and 1985 were not analyzed). The source of the discrepancies should be identified and 

resolved prior to the final calculation of coworker intakes. 

 

Finding 11: It is important to understand the reasons for the drastic decrease in observed 

sampling for uranium from 1984 to 1985. Concurrently, because the paucity of the data makes it 

impossible to make quantitative statements about the available bioassay data from 1984 to1985, 

it should be established that conditions and exposure potential are sufficiently similar to (or 

bounded by) surrounding years to validate the use as surrogate data. 

Finding 12: Uranium intake rates appear to be underestimated, based on the practice of 

collecting urinalysis samples on Monday morning after 2 days of no exposure. Although not all 

the available bioassay data were collected on Monday morning, NIOSH should apply an 

adjustment factor to the calculated intakes as a claimant-favorable and bounding assumption. 

Finding 13: NIOSH should discuss how uranium samples taken from workers in the uranium 

aluminide (U-Alx) powder operations are interpreted in the coworker analysis, and whether a 

separate assessment of coworker doses in the Powder Building may be warranted. In the latter 

case, it would be necessary to establish with sufficient accuracy which workers were potentially 

exposed to U-Alx in the Powder Building during the relevant period.

Finding 14: The internal coworker model should specify the types of incidents that are assumed 

to be covered by the model, and when incidents would be expected to be reviewed separately 

with an incident-specific model. 

Finding 15: The internal dose coworker model is incomplete in that it does not address other 

radionuclides that were present on site and that could have substantively contributed to worker 

intakes. NIOSH should ascertain their ability to reconstruct intakes of thorium, tritium, 

americium, cesium, strontium, polonium, and cobalt. 
Effective Date:

November 24, 2014

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Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

In addition, SC&A identified the following 10 pertinent observations: 

Observation 1: The assumption that database entries that do not contain a specific sample 

volume and are in units of total activity are reflective of a daily excretion rate has been validated 

by examination of handwritten logbooks contained in individual claimant bioassay records. 

Observation 2: SC&A noted that when the “reported” value was not designated as “<”, the 

difference between the reported and calculated values was simply that the “reported” result was 

normalized to 1,500 mL. Since NIOSH is already making the volume correction for the 

“calculated” result, this has a negligible effect on the coworker calculations. 

Observation 3: SC&A noted that the value reported in the “Remarks” box was not always 

transferred to the ORAUT-created electronic database. NIOSH needs to ensure that all reported 

doses in the hardcopy of the dose records are appropriately transferred to the electronic database. 

Observation 4: To better understand what the overall monitoring data represent, research 

should be performed to determine why no monitoring data are available for certain occupations 

in the late 1980s and early 1990s. 

Observation 5: Page 9 of ORAUT-OTIB-0080 states, “Sufficient plutonium bioassay data were 

available to perform a statistical analysis for 1965 through 1986.” Page 11 of ORAUT-OTIB0080 states, “Insufficient [plutonium] bioassay data were available for 1965 and 1966 to perform 

a statistical analysis... The 1967–1968 data was used to back-extrapolate for 1965 and 1966, 

when the potential plutonium exposure was less.” It appears based on the analysis and 

discussion performed that the latter statement is correct and that the former statement should be 

revised. 

Observation 6: The applicable plutonium data contained two results with units of weight 

(2.69 gm and 0 μg), both marked as “urine” results. Both results were set to zero dpm/1.4L. In 

the case of the former result, it is unclear why the sample was set to zero; however, the effect of 

this adjustment is likely negligible. Regardless, ORAUT-OTIB-0080 specifies on page 8 that 

samples given in units of mass are assumed to be fecal samples and are not to be used. However, 

these two samples were marked as “Y” in the “Use” column of the database. 

Observation 7: Bioassay data in available literature pertaining to the monitoring of plutonium 

intakes support the assumption made in the internal dose model that the intake rate in the late 

1980s and 1990 is representative (or even bounding) of plutonium intakes that could have 

occurred in the 1990s. 

Observation 8: While there are temporal gaps in the job title coverage for mechanics and 

technical staff, less than 1⁄4 of the monitored workers could be identified with a particular 

occupation. Of the monitored workers for which job title information exists, mechanics and 

technical staff constitute 65% of the population. Therefore, it is reasonable to assume that these 

higher risk job types were not systematically excluded and were likely targeted for plutonium 

monitoring. 
Effective Date:

November 24, 2014

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However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

Observation 9: Note that the last effective bioassay date in Table 15 is March 1, 1991. 

Examination of the available data indicates that “FP” bioassay data effectively end in mid-April. 

This comports with an effective bioassay date for the evaluation period of January 1, 1991, to 

roughly April 30, 1991. Since the end of April 1991 marks the end of the “FP” bioassay 

evaluation, the calculated intakes found in ORAUT-OTIB-0080 Table 5-6 should also end in 

April 1991 and not December 1991. However, it is not clear if changing the date of the end of 

the chronic intake calculation would significantly affect the calculated intakes. 

Observation 10: It is important that assumptions that cross-cut among models (e.g., internal, 

external, ambient) are consistent. Unless there is justification to do otherwise, the length of time 

of exposure of coworkers should be assumed to be the same across all models. Because use of 

overtime is often a common practice among operations personnel, a 2,000 work-year may not 

bound the dose to a worker. 
Effective Date:

November 24, 2014

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Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

1.0 CHARACTERIZATION OF AVAILABLE COWORKER DATASET 

SC&A requested and was provided with the underlying dataset containing the internal 

monitoring data utilized in ORAUT-OTIB-0080 to calculate coworker intake rates.2 The 

database contains 37,323 total entries covering various types of internal monitoring, including 

urinalysis, whole body counts, wound monitoring, fecal results, and smear samples. The 

available internal monitoring results also cover a myriad of different contaminants and radiation 

types. In the database, 118 distinct “Nuclide/analysis” types were identified, although several of 

these entries were redundant, such as “GB,” “Beta,” Gross Beta.” For a full listing of internal 

monitoring entries and the number of samples associated with each, refer to Attachment 1. 

2

 This dataset can be found at [K:\ABRWH\AB Document Review\Santa Susana\SSFL Internal coworker 

data for OTIB-0080 Rev 00\SSFL_FINAL compilation 0301113_r2 internal only R3.xlsx]. 

Many of the database entries were not germane towards calculating coworker intake rates for 

uranium, plutonium, and mixed fission products, and therefore had to be removed. The specific 

steps that SC&A utilized to remove unwanted samples are documented in Table 1 and comport 

with the analysis steps discussed in ORAUT-OTIB-0080 and Arno 2013. As seen in Table 1, 

after the steps to remove unwanted internal monitoring entries were performed, there were 902 

plutonium results, 3,378 mixed fission product results, and 8,669 uranium results available for 

coworker analysis. Each of these sub datasets is evaluated and discussed in further detail in 

Sections 3-5, respectively. 

Table 1. Description and Rationale for the Removal or Extraction of Entries from the 

Database for Coworker Analysis

Total Entries 

Remaining for 

Analysis

Rationale for Items Removed/Extracted

37,322 No entries removed; this represents the total number entries in the original database.

26,515

Remove all entries marked as "N" in the "Use" column (Column AU) indicating the sample 

should not be used in coworker analysis. Remove samples prior to 1/1/1965 or after 8/4/1991. 

The end date was chosen to mirror coworker modeling dates and exclude samples analyzed by 

the laboratory “Controls for Environmental Pollution, Inc.” (CEP) which has been determined 

to be an unreliable bioassay vendor. 

25,546

Remove entries with a blank field in the “Nuclide/Analysis” column (column T). Remove 

entries with a blank result in both the “Reported result” and “Calculated result” columns 

(columns AC and AD, respectively) in accordance with Arno 2013. 

902

Extraction of Plutonium Coworker Records: Extract all records with “Nuclide/Analysis Type” 

PU, Pu and PUA in accordance with ORAUT-OTIB-0080, pg. 9 and Arno 2013, pg. 2. 

Remove two additional records with in vitro units of “gm” (grams) and “μg” (micrograms).

3,378

Extraction of Fission Product Coworker Records: Extract all records with “Nuclide/Analysis 

Type: MFP” with a corresponding “Method Type: B.”3 Additionally, extract all records with 

“Nuclide/Analysis Type: FP” which contains a corresponding “Method Type: 3A.”4 The 

selected bioassay data are in accordance with ORAUT-OTIB-0080, pg. 10.

8,682

Extraction of Uranium Coworker Records: Extract all records with “Nuclide/Analysis Type: 

UR” in accordance with ORAUT-OTIB-0080, pg. 9 and Arno 2013, pg. 2. Remove two 

additional records with in vitro unit types “μg” (micrograms) and “UR” (unknown unit). 

 

3

 This includes “Nuclide/Analysis” entries “MFPB” and “MFP(B).” Additionally MFP with method types: 

“1B”, “2B”, “3B”,”5B”, “7B” 

4

 This includes method types: “3A T-U” and “3AG” 
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Each datapoint available for coworker analysis was normalized to the units of dpm/1.4L. SC&A 

noted that there was a large number of activity units used in the coworker data. Table 2 presents 

the distribution of observed units for each of the coworker radionuclides of interest. As seen in 

the table, the majority of samples for each monitored radionuclide were given in the units of 

“dpm.” In a small percentage of cases where the unit “dpm” was observed, there was a 

corresponding sample volume, which can be used to normalize the result to dpm/1.4L. For 

example, only 22 of 666 plutonium samples, 11 of 2,611 mixed fission product samples, and 810 

of the 4,986 uranium samples given in “dpm” contained a corresponding sample volume. The 

remaining samples given in units of dpm did not state the volume of the sample. 

Table 2. Overview of Bioassay Sample Units Available for Coworker Analysis 

Observed Units

# Plutonium 

Samples 

(% of Total)

# Fission Product 

Samples

(% of Total)

Uranium

dpm 666 (73.84%) 2611 (77.29%) 4986 (57.52%)

dpm/sample 62 (6.87%) 266 (7.87%) 2925 (33.74%)

dpm/L 10 (1.11%) Not Used Not Used

dpm/d 133 (14.75%) 221 (6.54%) 41 (0.47%)

dpm/vol 2 (0.22%) 14 (0.41%) 4 (0.05%)

Blank 29 (3.22%) 266 (7.87%) 712 (8.21%)

cpm Not Used Not Used 1 (0.01%)

Also as shown in Table 2, other common units observed in the database were actually provided 

in dpm per unit volume (such as dpm/d or dpm/L). With the exception of the samples in which a 

separate sample volume was given or the units were already provided in “dpm/d” and “dpm/L,” a 

sample volume would seemingly be required to correctly normalize the numerical result to 

dpm/1.4L. SC&A examined each coworker dataset to determine how many results provided 

enough information to effectively normalize the sample. The results are shown in Table 3.5

5

 dpm/d and dpm/L were assumed to represent sample volumes of 1.5 and 1 liter, respectively. 

Table 3. Percentage of Entries for Each Contaminant with Known and 

Unknown Sample Volumes 

Contaminant 

Type

Total Entries 

Available for 

Coworker Analysis

Percentage of 

Samples with a 

Known Volume*

Percentage of 

Samples with an 

Unknown Volume*

Plutonium 902 21.51% 78.49%

Fission Products 3,378 15.04% 84.96%

Uranium 8,669 39.75% 60.25%

*Samples listed as dpm/d or dpm/L were assumed to be 1.5 liters and 1 liter respectively 

As can be seen in Table 3, the majority of samples in each contaminant category did not contain 

the volume of the sample analyzed. In the case of fission products, nearly 85% of the samples 

available for coworker analysis did not have an associated sample volume. The sample volume 

of a particular aliquot given only in units of total activity can have a significant effect on the 

normalized end result; therefore, it is essential that the volume of these results be understood. 

 
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ORAUT-OTIB-0080 states the following on page 9 concerning unknown sample volumes: 

 

Some of the records have individual fields that are blank or illegible but contain 

sufficient information to proceed with the analysis. The following rules were used 

to adjust for blank or unusable fields: ... Volumes were assumed to be 1,500 ml. 

Additionally, Arno 2013 provides similar instructions: 

 Uranium: “If no volume is provided, assume 1500 mL.” (pg. 2) 

 Plutonium: “If no volume is provided, with units of dpm or dpm/sample, assume results 

are dpm/day.” (pg. 2) 

 Fission Products: “If no volume is provided, assume 1500 mL.” (pg. 3) 

The technical justification for the statements in ORAUT-OTIB-0080 was annotated as follows: 

Arno, Matthew. ORAU Team. Dose Reconstructor. January 21, 2011. These 

rules were developed based on examination of the data for trends in how the data 

was recorded, review of the internal dose technical basis document for SSFL 

(ORAUT 2010) to determine default, and/or assumed analytical techniques and 

reporting conventions.

The Technical Basis Document (TBD) for Occupational Internal Dose, ORAUT-TKBS-0038-5

(ORAUT 2010) contains the following statements about the bioassay program in the 1967–1974 

period: 

 The detection limit [for Uranium Radiometric (UR)] was 0.5 dpm/sample. At 

SSFL the standard sample volume per day was 1,500 mL. The result was the total 

activity.” (pg. 19) 

 The detection limit [for plutonium] was 0.05 dpm/sample. At SSFL the standard 

sample volume per day was 1,500 mL. (pg. 19) 

No direct references appear to be provided in ORAUT 2010 for the above two statements and no 

mention of standard sample volumes was provided for mixed fission products. Information on 

the standard sample volume was not stated before or after the 1967–1974 period in the TBD. 

SC&A examined the data available for coworker analysis to assess the extent to which the data 

itself lend credence to the assumption that the standard samples were 1.5 liters. The percentage 

of records that list a corresponding sample volume are shown by year in Figure 1. 
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Figure 1. Percentage of Records Listing a Sample Volume for Uranium, 

Plutonium and Fission Products 

As seen in Figure 1, it is clear that the majority of results for uranium and plutonium contained a 

listed sample volume in the 1965–1967 timeframe. After this time, generally 90% of the 

available samples for uranium and plutonium did not contain a sample volume. One possible 

conclusion for plutonium and uranium is that after 1967, the actual sample volume was not 

recorded, but was taken into account in the reported result. Another possibility is that the actual 

sample volumes were simply not entered into the available database. There does not appear to 

be a discernible trend for fission product bioassay results. 

In order to verify the assumption that samples given in units of total activity were, in fact, 

representative of daily excretion rates, SC&A examined the individual claimant monitoring 

records contained in the NIOSH/OCAS Claims Tracking System (NOCTS) database for 

additional information to affirm or refute the assumed sample volumes. SC&A found several 

cases where handwritten logbooks were included in the DOE-supplied monitoring records that 

indicated the sample was in the units of “dpm/day” (nominally dpm/1,500 ml). These same 

numerical values could be matched up to the electronic record, which either contained no units or 

contained units of total activity (dpm). One example of this can be seen in Figures 2 and 3, 

which show the handwritten log and associated database entry. 
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gure 2. Example Logbook Contained in Claimant Hardcopy File Showing Units of 

“dpm/day”

Fi

Figure 3. Corresponding Entry in Coworker Database: Entries Showing Units of “dpm” 

with a Blank Field for “Analyzed Volume”

Observation 1: The assumption that database entries that do not contain a specific sample 

volume and are in units of total activity are reflective of a daily excretion rate has been validated 

by examination of handwritten logbooks contained in individual claimant bioassay records. 
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Finally, as seen in Figure 4, there are two sources of urinalysis activity values present in the 

SSFL records (see columns AC and AD in the figure). The column AC header is titled 

“Reported Result (in box),” while the column AD header is titled “Calculated Result (above 

box).” In some cases there is a numerical value for each entry, but in most cases a numerical 

value is present in only one of the columns. 

Figure 4. Example Database Entry Showing “Reported” and “Calculated” Urinalysis 

Activity Results 

Arno 2013 instructs the modeler to always use the “calculated result” when present and only to 

use the “reported result” when no calculated result is available. Appendix B of ORAUT 2010 

provides descriptions of the different bioassay forms encountered at Santa Susana. An 

illustrative example of a bioassay form containing both a calculated and reported value can be 

found in Figure 5. 

The figure highlights the formula used to reach the calculated value, as well as the reported value 

and sample volume. Notably, the reported value is quantitatively higher and is labelled as a 

censored result (i.e., “less than” or “<”). The reported value is likely indicative of the MDA for 

the given measurement and sample volume. 
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Figure 5. Example Bioassay Form Showing Both a “Calculated” Value and 

a “Reported” Value

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The uranium coworker data contained 1,719 entries that listed a numerical value for both the 

“reported” and “calculated” entries (roughly 20% of the total coworker samples). SC&A closely 

examined these paired results for evidence of a numerical relationship and it appears these values 

are in fact MDA values that have been prorated to the specific volume of the sample. Figure 6 

plots the sample volume versus the reported result for samples labelled with “<.” While there is 

some scatter present in the figure, there certainly appears to be a linear relationship between the 

“reported” value and the sample volume.

Figure 6. Less than “Reported” Results versus the Associated Sample Volume 
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Additionally, Table 4 characterizes the “reported” result and “calculated” result quantities for the 

most commonly observed sample volumes.6

 Column 2 of the table shows that approximately 

78% of the paired samples for which the “reported” result was censored fell into one of the three 

stated sample volumes. Columns 3 and 4 show the most common “reported” value by sample 

volume as well as the percentage of entries reflecting that value. Columns 5–7 characterize the 

spread of “calculated” values for that same sample. As an example, of the 135 samples of 

90 mL, 96% of the “reported” values were <0.54 dpm. The “calculated” result for these same 

samples ranged from 0 to 0.52 with an average of 0.14. It is clear from Table 4 that the reported 

values have been truncated at a censoring level, while the “calculated” values span a wide range 

of values below this censorship level. 

6

 There were 80 different sample volumes present in the database for paired results with a “<” reported 

result. The three most common were 90 ml (~8% of the entries, 95 ml (~39% of the entries), and 100 ml (~31% of 

the entries). 

Table 4. Characterization of Paired Results in which the “Reported” Value was 

Designated as “Less Than”

Sample 

Volume 

(mL)

# of Total 

Entries (% 

of Total)

Most 

Common 

“Reported” 

Value 

(dpm)

% of “Reported” 

Samples Exhibiting 

Most Common 

Value for Indicated 

Sample Volume

Minimum 

“Calculated”

Value

Average 

“Calculated” 

Value

Maximum 

“Calculated”

Value

90 135 (8%) <0.54 96% 0 0.14 0.52

95 642 (39%) <0.57 99% 0 0.17 0.6*

100 508 (31%) <0.60 98% 0 0.20 0.8**

*Only 1 of the 642 “calculated” values exceeded the “reported” result of 0.57. 

**Only 3 of the 508 “calculated” values exceeded the “reported” result of 0.60. 

Finding 1: It is evident that the “reported” values with the “<” designation represent the 

volume-adjusted MDA and that the “calculated” values are much less than the censorship level. 

In these instances, NIOSH should utilize the “reported” value as a censored result, which is 

claimant favorable and in accordance with the MPM methodology in ORAUT-RPRT-0053 

(ORAUT 2012).

Observation 2: SC&A noted that when the “reported” value was not designated as “<,” the 

difference between the reported and calculated values was simply that the “reported” result was 

normalized to 1,500 mL. Since NIOSH is already making the volume correction for the 

“calculated” result, this has a negligible effect on the coworker calculations. 

Observation 3: SC&A noted that the value reported in the “Remarks” box was not always 

transferred to the ORAUT-created electronic database (for example, see the dose records for 

claimant ). NIOSH needs to ensure that all reported doses in the hardcopy of the dose 

records are appropriately transferred to the electronic database.

 
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2.0 OVERVIEW OF MONITORED JOB TYPES CONTAINED IN 

ORAUT-OTIB-0080 DATABASE

The electronic dose record database prepared by ORAUT was examined to determine what 

occupations were primarily monitored. The database contained a data field titled ‘Job Title.’ 

This data field contained the job title of an individual contained on their hardcopy dose record if 

it was identified and legible. Of the 37,323 dose records in the database, 7,152 (19%) had job 

titles identified. To increase the number of records with identifiable job titles, the NOCTS 

database was compared to the dose records database, and job titles (occupational names) were 

added to the electronic database based on individuals’ social security numbers. If the job title 

identified in the dose record database was different than the job title identified in NOCTS, then 

the NOCTS database replaced the dose record database job title information. In addition, for 

some individuals in NOCTS, their occupation changed during the course of their career. This 

change in occupational status was reflected in the dose record database such that an individual 

may be counted more than once when totaling the number of individuals monitored for each 

occupation. In addition, because many different job titles were used, for the purposes of this 

analysis, job titles were grouped into more generalized occupational titles. Attachment 3 shows 

how different job titles were grouped together. 

The addition of information from NOCTS increased the number of dose records with an 

identifiable occupation to 9,149 records. Records prior to 1965 and between January 1991 and 

mid-June 1993 were removed, per the procedure applied in the internal dose model, which 

reduced the total number of records to 8,856. Deleting records with the ‘Use’ designation of ‘N’ 

reduced the number of records to 6,550. Removing all records with no ‘Nuclide/Analysis’ 

identified and no ‘Reported Result’ and ‘Calculated Result’ reduced the number of records to 

6,300.

Table 5 shows the results of the occupational analysis. The most prevalent occupations that were 

monitored were technical staff (30% of the monitored individuals) and mechanics (26% of the 

monitored individuals). This would be expected, because the mechanics would have performed 

the majority of the hands-on activities associated with nuclear operations with the support of the 

engineers and scientists who comprise the technical staff and who would have been responsible 

for the design, construction, and operation. The cross-section of occupations represented by the 

dose records also includes occupations not directly related to performance of nuclear operations, 

but who may have been periodically present. These support personnel represent occupations 

such as management, office personnel, security, and inspectors, as shown in Table 5. The 

distribution of workers shown in Table 5 appears to be representative of a monitoring program 

that focuses on those occupations at highest risk of exposure (technical staff and mechanics), but 

also reflective of support personnel in the area. 

Note that there are seven firemen that were monitored, all of whom had recorded dose in at least 

one monitoring period. Survivors of workers stated in the February 9, 2010, Advisory Board 

meeting that they were concerned that firemen would not be compensated, because there was no 

proof they worked in Area IV. While it is not possible to verify the presence of specific 

individual firemen in Area IV unless they are shown in the electronic database, this occupational 

analysis shows that at least some firemen were monitored at both DeSoto and Area IV. The 
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occupational doses that these monitored firemen received would be included in the internal 

coworker model, though the question of how well the available data represent or bound intakes 

received by unmonitored firemen remains open. 

SC&A notes that monitoring data for many of the occupations end in 1982–1983. This can be 

expected for an occupation such as Reactor Operator, because the last operating reactor ceased 

operations in 1980. However, it is unclear why other occupations, such as Health Physics and 

Maintenance, do not have monitoring data continuing into the early 1990s, consistent with 

several other occupations, despite the continued operation of other SSFL radiological facilities, 

such as the Hot Lab Facility (operational through 1988), the Fuel Storage Facility (operational 

through 1993) and the Radioactive Materials Handling Facility (currently operational). For the 

two most heavily monitored occupations (Technical Staff and Mechanics), there are data 

available for every year through 1991, although not necessarily data for every contaminant in 

each year (see Sections 3.4, 4.3 and 5.4). 

Observation 4: To better understand what the overall monitoring data represent, research 

should be performed to determine why no monitoring data are available for certain occupations 

in the late 1980s and early 1990s. 
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Table 5. Occupations Monitored at SSFLa

Occupational Group Number of 

Records Range of Years Number of Individualsb

Percentage of Individuals 

in an Occupational Group

Construction 25 1965–1997 [redacted] 2%

Electrician 131 1965–1988 10 3%

Fireman 33 1965–1974 [redacted] 2%

Health and Safety/

Health Physicsc

281 1964–1988 12 4%

Inspector 38 1975–1994 [redacted] 1%

Janitor 310 1965–1983 12 4%

Machinist 112 1965–1983 [redacted] 3%

Maintenanced203 1965–1983 17 5%

Management 208 1965–1986 [redacted] 2%

Mechanice2,446 1963–1993 81 26%

Office Personnel 153 1965–1989 12 4%

Operator 58 1965–1976 [redacted] 2%

Quality Assurance 63 1970–1986 [redacted] 1%

Radiographer 24 1975–1982 [redacted] <1%

Reactor Operator 85 1965–1983 [redacted] 2%

Security 51 1965–1989 [redacted] 3%

Shipping 8 1965–1968 [redacted] <1%

Technical Stafff 1,531 1965–1993 96 30%

Technician 185 1949 –1997 [redacted] 3%

Truck Driver 86 1966–1986 [redacted] 1%

Unknowng241 1965–1981 [redacted] 1%

Welder 51 1965–1982 [redacted] 1%

Total 6,323 - 316 -

a

 Excludes dose records prior to 1965 and between January 1991 and mid-June 1993.

b

 An individual may be represented more than once if their occupation changed over the course of their 

monitoring.

c

 Includes medical staff, and health physicists and their support staff

d

 Includes individuals with ‘maintenance’ as part of their job title, utility services staff, and painters

e

 Includes individuals with ‘mechanic’ somewhere in their job title, and those who describe themselves as reactor 

and reactor fuel assemblers/disassemblers.

f

 Includes individuals with ‘engineer’ as part of their job title, physicists, chemists, and analysts

g

 Either a job title was not specified and marked as unknown, or it is unclear what the job title represents (for 

example, ‘Y&S’, ‘FMS’).

More specific occupational information related to plutonium, fission product, and uranium 

monitoring is provided in Sections 3.4, 4.3 and 5.4, respectively, of this report. 
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3.0 REVIEW OF PLUTONIUM DATA AND ASSOCIATED 

COWORKER INTAKES 

Section 3 provides a detailed review of the plutonium coworker model presented in ORAUTOTIB-0080. Specifically, the OPOS values are examined for applicability and appropriateness 

in coworker modeling (Section 3.1). The raw data were also used to try and recreate the OPOS 

results presented in Attachment 1 of ORAUT-OTIB-0080 (see Section 3.2 below). A discussion 

of plutonium operations throughout the period of interest can be found in Section 3.3. 

Section 3.4 presents an analysis of the occupations monitored for internal plutonium exposures. 

3.1 REVIEW OF CALCULATED OPOS EXCRETION RESULTS 

Calculated urinary excretion rates (OPOS values) from ORAUT-OTIB-0080 are presented in 

Table 6. 

 

Table 6. Plutonium Excretion Rates (dpm/1.4L) 

Effective Bioassay Date 50th Percentile 84th Percentile No. of employees

1/1/1968 0.064 0.328 34

1/1/1970 0.026 0.097 31

1/1/1974 0.013 0.040 28

1/1/1976 0.037 0.049 48

1/1/1980 0.022 0.034 30

1/1/1984 0.040 0.053 48

1/1/1986 0.020 0.031 53

 Source: ORAUT-OTIB-0080, Attachment A, Table A-1 

As seen in Table 6, the calculated 50th and 84th percentile OPOS values were not analyzed over 

the standard period of 1 year, as can be seen in the choice of effective bioassay dates (column 1). 

The effective bioassay date represents the midpoint of a given OPOS evaluation period. For 

example, if the evaluation period was the standard “1 year” period, the effective bioassay date 

for that period would be on July 1st (approximately halfway through the evaluation period). In 

the case of plutonium, NIOSH did not begin to use the plutonium data until 1967; specifically, 

ORAUT-OTIB-0080 states: 

Insufficient [plutonium] bioassay data were available for 1965 and 1966 to 

perform a statistical analysis... The 1967–1968 data was used to back-extrapolate 

for 1965 and 1966, when the potential plutonium exposure was less.

Therefore, it can be assumed that the first OPOS evaluation period began on January 1, 

1967, with its midpoint on January 1, 1968, and its endpoint on December 31, 1968 (or 

alternately January 1, 1969). Similarly, the OPOS evaluation of available plutonium data 

ended in 1986 and has an “effective bioassay date” of January 1, 1986. Therefore, the 

final evaluation period would have started on January 1, 1985. Using this logic, SC&A 

attempted to match the effective bioassay dates shown above in Table 6 with the period 

under evaluation. Table 7 demonstrates this attempt to match up the effective bioassay 

dates with appropriate analysis intervals. 
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Table 7. Plutonium OPOS Analysis Period Reconstruction Based on Effective Bioassay 

Dates Shown in ORAUT-OTIB-0080 Table A-1 

ORAUT-OTIB-0080 Effective 

Bioassay Date

SC&A Assumed 

Start Date

SC&A Assumed 

End Date

Years Included 

in OPOS Result

1/1/1968 1/1/1967 12/31/1968 2

1/1/1970 1/1/1969 12/31/1970 2

1/1/1974 1/1/1971 12/31/1976 6

1/1/1976 Unknown Unknown -

1/1/1980 1/1/1977 12/31/1982 6

1/1/1984 1/1/1983 12/31/1984 2

1/1/1986 1/1/1985 12/31/1986 2

The method SC&A used in Table 7 is the same as the example described above. The 

assumptions were that the starting date for the first acceptable plutonium analysis period was 

January 1, 1967, and the ending date for the final acceptable plutonium analysis period was 

December 31, 1986. As one can see in the table, the middle evaluation period highlighted in red 

does not comport with the surrounding evaluation periods. In fact, the effective bioassay date of 

January 1, 1976, appears to be covered by the previous evaluation period, which spans from 

January 1, 1971, to December 31, 1976. If the highlighted entry in Table 7 was removed, then 

the remaining evaluation periods and effective bioassay dates cover the entire period seamlessly. 

Finding 2: SC&A was unable to reconcile apparent differences between the effective bioassay 

dates listed in Table A-1 of ORAUT-OTIB-0080 and the known start and end dates of 

“statistically usable” plutonium data. 

Additionally, it should be noted that whatever the correct evaluation periods should be, there will 

necessarily be evaluation periods that cover more than 1 year. It is assumed that these OPOS 

evaluation periods were chosen to reach the recommended “30 workers” per period in 

accordance with ORAUT-RPRT-0053, Analysis of Stratified Coworker Datasets (ORAUT 

2012). RPRT-0053, however, also states: 

Data from multiple years (usually no more than 3) can be combined to achieve 

this minimum if the conditions in the workplace are reasonably constant over the 

period in question. (ORAUT 2012, pg. 9) 

Similarly, SC&A found in its review of that report that combining periods of bioassay for the 

purposes of calculating OPOS values must be validated and cannot be based solely on gaining 

the desired number of samples. It must be demonstrated that the combination of longer periods 

is appropriate and defensible from an exposure potential standpoint. On the other hand, it has 

been discussed at recent SEC Work Group meetings that a given evaluation period in which less 

than 30 OPOS results are available may be acceptable if the exposed population was very small 

(i.e., bench scale operations in which one would expect a low number of monitored workers). 

Neton 2014a has the following statements regarding the requirements of sample sizes for 

coworker analysis: 
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The amount of the available monitoring data should also be evaluated to 

determine if there are sufficient measurements to ensure that the data are either 

bounding or representative of the exposure potential at the facility. Facilities 

with the potential for internal and/or external exposure to a large percentage of 

the workforce would require many more samples than one in which the potential 

for exposure was limited to just a few workers. In addition, the variability of the 

exposure potential should be considered. It has been observed, for example, that 

some National Laboratories conducted work under many different experimental 

configurations, resulting in a wide variety of exposure potentials. In this case, it 

might not be possible to generate a single coworker model that adequately 

captures all categories of unmonitored worker doses. 

 

Although there is no hard and fast rule for the minimum number of data points 

required to represent a given time interval, approximately fifteen values has [sic] 

been cited as a reasonable number for performing statistical tests on censored 

datasets (Singh et al. 2010). Because our program estimates parameters from the 

data, a default minimum of 30 values is recommended per modeled interval. The 

minimum number of samples should, of course, be considered in light of the 

number of worker’s [sic] potentially expose [sic] to the airborne source-term. 

(Neton 2014a, pg. 2) 

SC&A understands that these global policy discussions are ongoing and many post-date the 

release of this coworker model. However, these newer developments should be considered in the 

discussions and issues resolution process for this internal dose reconstruction method going 

forward. 

Finding 3: Combining multiple years for the purpose of calculating the 50th and 84th percentile 

OPOS results must be scientifically justified in the context of the potential for historical changes 

in operations and associated intakes for monitored workers. 

ORAUT-OTIB-0080 determined that there was not a sufficient number of plutonium bioassays 

in the first 2 years of evaluation (1965–1966) to perform statistical evaluation. As stated 

previously in this section, the rationale for back extrapolation was that plutonium exposure 

potential was less in 1965–1966 than it was in 1966–1967, and therefore the back extrapolation 

of intakes is claimant favorable. Specifically, ORAUT-OTIB-0080 (pg. 11) states: 

The Nuclear Materials development facility began operation in 1967, which led to 

an increase in potential plutonium exposure and therefore plutonium urinalysis 

data. The 1967-to-1968 data was used to back-extrapolate for 1965 and 1966, 

when the potential plutonium exposure was less. This assumption is favorable to 

the claimants.

SC&A agrees with NIOSH that the data are very limited during those first 2 years of potential 

evaluation, with only 14 bioassay samples identified. While these samples may not be useful for 

the standard statistical analysis, they do contain valuable information to indicate exposure 

potential during those years. SC&A performed a simple rank ordering of the available 1965–
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1966 data for comparison with the calculated OPOS excretion rates at the 50th and 84th percentile 

shown in Table A-1 of ORAUT-OTIB-0080. This comparison is shown in Figure 7. The blue 

dots represent the actual plutonium bioassay data for 1965–1966, the red dots represent the 

calculated excretion data from ORAUT-OTIB-0080, and the green dots represent the linear 

interpolation between the actual 1965–1966 data to obtain an approximate 50th and 84th

percentile value. 

Figure 7. Comparison of Available 1965 and 1966 Plutonium Urinalysis Values Including 

th th Interpolated Values at the 50 and 84 Percentile with Calculated OPOS Results from 

th th 1967–1968 at the 50 and 84 Percentile 

Though the data are limited, the calculated 50th and 84th percentile excretion rates shown for 

1967–1968 do not bound the rank-ordered data for the earlier period. In order to justify the back 

extrapolation of one period to earlier period, one must clearly demonstrate that the exposure 

potential is bounded by the later period. This can be accomplished through work place 

monitoring, BZ/daily weighted exposure studies, source term characterizations, etc. SC&A 

concedes that the increase in plutonium monitoring likely mirrors an increase in the number of 

workers exposed to plutonium. However, it cannot be assumed a priori that this indicates an 

increase in intake potential to the exposed worker. 

Finding 4: More definitive evidence is warranted to establish that exposure potential to 

plutonium was lower in 1965–1966 than it was in 1967–1968 to validate the back extrapolation 

of plutonium results to the earlier period. 
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SC&A also had the following observation pertaining to the text of ORAUT-OTIB-0080: 

Observation 5: Page 9 of ORAUT-OTIB-0080 states, “Sufficient plutonium bioassay data were 

available to perform a statistical analysis for 1965 through 1986.” Page 11 of ORAUT-OTIB0080 states, “Insufficient [plutonium] bioassay data were available for 1965 and 1966 to perform 

a statistical analysis... The 1967–1968 data was used to back-extrapolate for 1965 and 1966, 

when the potential plutonium exposure was less.” It appears based on the analysis and 

discussion performed that the latter statement is correct and that the former statement should be 

revised. 

3.2 REPLICATION OF CALCULATED PLUTONIUM OPOS VALUES 

SC&A was provided with the raw database used by ORAUT in calculation of plutonium OPOS 

values shown in Table A-1 of ORAUT-OTIB-0080. SC&A extracted the appropriate sample 

results based on the criteria described in Section 1 of this report. SC&A then calculated OPOS 

values based on the assumed evaluation period shown in Table 6 in Section 3.1. As was noted in 

that section, SC&A was not able to determine the exact evaluation periods utilized by ORAUT 

for the purposes of calculating OPOS, and there is some confusion as to how the middle 

evaluation periods were obtained. Nonetheless, it appears that comparison of the early 

evaluation periods and later evaluation periods has some merit based on the total number of 

monitored workers. Table 8 contains the results of SC&A’s OPOS calculation as compared to 

the ORAUT Table A-1 values. 

Table 8. Comparison of SC&A-Calculated OPOS Results and ORAUT-OTIB-0080 

Table A-1 OPOS Results 

Effective 

Bioassay 

Date

SC&A 

Assumed Start 

Date

SC&A 

Assumed 

End Date

# SC&A 

OPOS

# NIOSH 

OPOS

SC&A 50th

Percentile 

OPOS 

(dpm/1.4L)

NIOSH 50th

Percentile 

OPOS 

(dpm/1.4L)

SC&A 84th

Percentile 

OPOS 

(dpm/1.4L)

NIOSH 84th

Percentile 

OPOS 

(dpm/1.4L)

1/1/1968 1/1/1967 12/31/1968 33 34 0.0922 0.064 0.585 0.328

1/1/1970 1/1/1969 12/31/1970 31 31 0.0691 0.026 0.207 0.097

1/1/1974 1/1/1971 12/31/1976 77 28 0.0235 0.013 0.050 0.040

1/1/1976* N/A N/A N/A 48 N/A 0.037 N/A 0.049

1/1/1980 1/1/1977 12/31/1982 46 30 0.0234 0.022 0.049 0.034

1/1/1984 1/1/1983 12/31/1984 48 48 0.0277 0.04 0.047 0.053

1/1/1986 1/1/1985 12/31/1986 53 53 0.021 0.02 0.036 0.031

* SC&A did not calculate an OPOS result for this effective bioassay date entry because it appeared to fall within the 

assumed evaluation dates of surrounding entries (see Section 3.1). 

It is not surprising that NIOSH and SC&A obtained different OPOS results, considering the 

confusion concerning what evaluation periods were actually used. Other reasons might also 

account for the differences, such as data selection/extraction. Interestingly, the number of 

workers involved in the calculation of the OPOS for the first two periods and the last two periods 

were remarkably similar. However in several cases, the actual magnitude of the calculated 

excreta result was higher in SC&A’s calculation. The exact nature and source of these 

discrepancies is unknown, but should be resolved to assure that the calculated excreta values 

used in calculating coworker intakes are accurate. 
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Finding 5: SC&A was not able to duplicate the plutonium OPOS values presented in Table A-1 

to within a reasonable level of precision. The source of the discrepancies should be identified 

and resolved prior to the calculation of coworker intakes. 

SC&A also had the following observation concerning the use of the dose information in the 

electronic database: 

Observation 6: The applicable plutonium data contained two results with units of weight 

(2.69 gm and 0 μg), both marked as “urine” results. Both results were set to zero dpm/1.4L. In 

the case of the former result, it is unclear why the sample was set to zero; however, the effect of 

this adjustment is likely negligible. Regardless, ORAUT-OTIB-0080 specifies on page 8 that 

samples given in units of mass are assumed to be fecal samples and are not to be used. However, 

these two samples were marked as “Y” in the “Use” column of the database. 

3.3 DISCUSSION OF HISTORICAL PLUTONIUM PROCESSES 

To more fully explore whether Findings 3 and 4 are valid concerns, SC&A examined the 

operational history of plutonium processes at SSFL. 

Facility Operations 

Operations at SSFL have involved the use of plutonium radionuclides over much of the site’s 

operational period. The most prominent facility that used plutonium was the Nuclear Materials 

Development Facility (NMDF), which operated from 1967 through 1979. The NMDF was 

constructed specifically to work with plutonium. Its operating history is summarized in Table 9. 

Based on this table, it appears that process campaigns typically did not extend for more than 1 or 

2 years, which does not agree with the more lengthy OPOS periods assumed by NIOSH in 

Table 7 of this report, which SC&A attempted to replicate in Table 8. 

Table 9. Operations at the Nuclear Materials Development Facility 

Operating Period Operation 

Operating Period Operation

1967–1968 Development of Analysis Technologies for uranium-plutonium oxide fuels

4/1968–6/1969 Recycle of scrap uranium-plutonium fuel

7/1968–6/1970 Development of technologies to mix tungsten into uranium-plutonium carbide fuel

4/1970–9/1970 Preparation of samples for uranium-plutonium oxide irradiation studies

9/1970–3/1974 Idle

1974–1975 Bench scale tests-recovery of plutonium from simulated waste

1975 – 5/1977 Mixed uranium-plutonium carbide fuel fabrication

5/1977–11/1978 Partial decontamination and clean-up

11/1978–11/1979 Fabrication of depleted uranium carbide fuel

11/1979–10/1982 Idle

10/1982–10/1986 Decontamination and decommissioning

7/1987 Released for unrestricted use

Source: Rucker 2009

Although plutonium fuel was manufactured at SSFL Area IV, according to operational 

descriptions, none of the reactors or criticality experiments operated at SSFL Area IV used 
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plutonium as a fuel. The manufactured plutonium fuel was actually a mixture of plutonium and 

uranium oxide pellets. The plutonium would have been comprised primarily of the isotope 

Pu-239, with lesser amounts of Pu-238, Pu-240, and Pu-241 (Rucker 2009). 

The following processes were used in about 1975 to 1977 (Rockwell 1975) and do not appear to 

be representative of earlier operations based on the description of activities in Table 9 of this 

report: 

1. Receipt, sampling, analysis, and storage of raw feed materials, including ceramic-grade 

PuO2 powder, enriched UO2 powder, and depleted UO2 powder 

2. Batch weighing, blending, and agglomeration of PuO2 powder, UO2 powder, and 

graphite 

3. Carbothermic reduction of oxide powders and graphite to plutonium-uranium carbide 

4. Crushing, milling and agglomeration

5. Chemical analysis of carbide sample 

6. Pressing of carbide into pellets 

7. Sintering of carbide pellets 

8. Product analysis 

9. Chemical analysis of selected pellets 

10. Metallographic examination of selected pellets 

11. Pellet dimensional and weight measurements 

12. Batching pellets for fuel pin loading 

13. Load pellets and pin hardware into fuel pins 

14. Settle fuel pellets and hardware into sodium 

15. Weld fuel pin end closure 

16. Centrifugally bond fuel pins

17. Fuel Pin Inspection 

18. Eddy current inspection 

19. Alpha monitor 

20. Leak test 

21. Autoradiograph 

22. X-ray welds and pin internal hardware 

23. Gamma scan 

24. Dimensional

25. Penetrate Xenon capsule

26. Package and ship fuel pins

27. Prepare, package, and ship scrap fuel material to recovery vendor 

The Hot Lab (Building 4020), which operated from 1959 to 1990, was used to examine reactor 

fuel assemblies and other test specimens to determine how they were performing. This involved 

handling and examining highly radioactive items. This facility was also used to support 

disassembly, fuel removal, and size reduction of items generated when a reactor was 
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decommissioned. In addition, the Hot Lab was used to manufacture and leak-check sealed 

sources, and machining operations involving Co-60. As a result of this work, the interior of the 

Hot Cells in the lab and associated equipment became contaminated with uranium, plutonium, 

thorium, and fission and activation products (ORAUT 2006a). No information was found in the 

Site Research Database (SRDB) that would indicate how operations may have changed over time 

in the Hot Lab. 

The Hot Cave in the Engineering Test Building (Building 4003) supported testing of nuclear fuel 

reprocessing from SSFL reactors and other off-site reactors as described in ORAUT-TKBS0038-2 (ORAUT 2006a), and operated from 1954 to 1964. The used fuel assemblies from 

nuclear reactors contain unused fissionable material and transuranic products, primarily 

consisting of plutonium. The facility was decontaminated and decommissioned in 1975 

(ORAUT 2006a).

Another facility associated with plutonium was the Fuel Storage Facility (Building 4064). This 

facility, which operated from 1958 to 1993, was used to provide safe, secure storage of fissile 

fuel material such as plutonium (Rucker 2009). 

Non-reactor-related research programs were performed at SSFL. One research program that 

required the use of plutonium was the TRUMP-S program, but this program was transferred to 

the University of Missouri. Seventy-five (75) grams of depleted uranium, 5 grams of plutonium, 

and 4 grams of neptunium, were received at SSFL Area IV and stored in the Fuel Storage 

Facility before being shipped to the University of Missouri sometime after 1990 (Rucker 2009). 

Plutonium Exposure Data 

SC&A reviewed available literature to determine if there were facility-specific or individualspecific plutonium exposure data available that would provide some indication of the prevalence 

and duration of plutonium-related activities. 

Some limited data are available that show how many NMDF workers received bioassays during 

short time periods in the early 1980s. Table 10 shows, by quarter, the number of people 

receiving bioassays, the number of samples taken, the number of analyses performed and number 

of positive results. No positive bioassay results were documented during these quarterly 

reporting periods. Based on Table 10, it appears that operations occurred intermittently, since no 

workers were monitored during some of the monitoring periods. 

Table 10. Bioassay Data for the NMDF 

Quarter Number of Individuals 

Monitored

Number of 

Samples

Number of Analyses 

Performed

Number of 

Positive Results

1980 Q2 3 3 4 0

1980 Q3 3 3 6 0

1980 Q4 0 0 0 0

1981 Q2 0 0 0 0

1982 Q2 3 3 8 0

1982 Q3 1 2 3 0

Source: Moore 1980a, 1980b, 1981a, 1981b, 1982a, 1982b 
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In another report (Barnes 2000), a screening analysis was performed to support regulatory 

compliance efforts at SSFL. In this study, a screening of all positive bioassay results between 

the years 1989 through 1999 was conducted. In the majority of cases, the screening 

demonstrated that Committed Effective Dose Equivalents (CEDEs) were less than 10 mrem. In 

certain cases, the screening results suggested that CEDEs greater than 100 mrem had been 

received. In these cases, formal dose assessments were conducted. Table 11 summarizes the 

results related to plutonium exposures. This report was the only one found that tied names of 

individuals to exposure results. The report did not identify the facilities in which these exposures 

took place. As can be seen, the exposures occurred over a short time span in 1989 and 1990, and 

did not extend through 1999. However, given that all documented plutonium activities had 

wound down by the late 1980s, as shown in Table 9, it is reasonable to assume that exposures 

monitored in 1989 and 1990 would have been higher than those in the mid to late 1990s due to 

generally accepted removal processes that deplete residual contamination. 

Table 11. Plutonium Bioassay Results from 1989–1999

Name Date of Intake Isotope (Class) CEDE (mrem) Method of 

Calculation

[redacted] 2/20/90 Pu-239 (W) 745 Assessment

[redacted] 12/19/89 Pu-239 (W) 45 Assessment

[redacted] 4/18/90 Pu-239 (W) <10 Assessment

[redacted] 7/14/90 Pu-239 (W) <10 Assessment

[redacted] 12/18/89 Pu-239 (W) 85 Screen

[redacted] 9/15/90 Pu-239 (W) 85 Screen

[redacted] 4/16/90 Pu-239 (W) 139 Assessment

[redacted] 7/14/90 Pu-239 (W) 26 Assessment

[redacted] 4/17/90 Pu-239 (W) 45 Screen

[redacted] 3/18/90 Pu-239 (W) <10 Assessment

[redacted] 5/14/90 Pu-239 (W) 284 Assessment

[redacted] 7/15/90 Pu-239 (W) 11 Assessment

Source: Barnes 2000 

Incidents Involving Plutonium 

The following two incidents were found in available literature in the SRDB that involved 

possible exposures to plutonium. No incidents were found in the DOE’s Occupational Reporting 

and Processing System, which contains incident reports since 1990. These incident descriptions 

provide a better understanding of the hazards that were encountered and protective measures 

used. 

On February 7, 1983, while disassembling a balance in glove box 3A in the 

NMDF, an employee accidentally jammed a small screwdriver through the box 

glove and surgeon's glove, resulting in a puncture wound at the base of his left 

index finger. He immediately notified Radiation and Nuclear Safety. A survey of 

the wound and finger indicated approximately 3500 dpm alpha. His surgeon's 

glove indicated approximately 30,000 dpm alpha. The total activity in the wound 

was estimated to be about 7 nCi. Surgery removed any remaining activity with 

bioassays showing negative results (Rockwell 1986). 
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On July 15, 1983, an incident involving irradiated plutonium fines and fission 

product radioactivity occurred at the Decon Room 4 alpha glove box in the Hot 

Lab. While preparing to replace a leaded glove with a plastic bag in a glove box, 

the glove box operator checked his gloved hands and detected alpha 

contamination. Personnel performed a survey that indicated gross alpha 

contamination on the hood, mask, lab coat of the operator (>50,000 c/m), as well 

as the glove port and the immediate surrounding area. Evacuation was ordered 

and additional support requested. Wearing respiratory protection, gloves, etc., 

two Radiation and Nuclear Safety representatives carefully stripped contaminated 

articles from the operator, leaving his full-face respirator until last. Activity 

levels up to 50,000 c/m alpha were detected, with 5,000 c/m alpha on the right 

lower part of the pant leg. The gross activity was removed (to less than 200 c/m 

alpha). His mask was then removed. Nasal smears of involved personnel 

indicated no significant radioactivity. Urine and fecal samples were requested 

and submitted for analyses. The urinalysis results were negative; the fecal 

specimen was lost at the laboratory during processing (Rockwell 1986). 

SC&A’s review of plutonium exposure data and incident descriptions leads to the following 

finding and observation: 

Finding 6: SC&A’s research into the historical use of plutonium at SSFL indicates that 

programs or campaigns in plutonium facilities were fairly diverse and often did not appear to 

extend past 1 or 2 years in most cases (at least in the case of the primary plutonium facility 

NMDF). Therefore, averaging plutonium data across multiple years for the purposes of 

calculating coworker intakes may not be appropriate. 

Observation 7: Bioassay data in available literature pertaining to the monitoring of plutonium 

intakes support the assumption made in the internal dose model that the intake rate in the late 

1980s and 1990 is representative (or even bounding) of plutonium intakes that could have 

occurred in the 1990s. 

3.4 OCCUPATIONS MONITORED FOR PLUTONIUM INTAKE 

Occupational information was analyzed by cross-referencing dose records that specify an 

analytical method for detecting plutonium with occupational information developed as described 

in Section 1.0. From Section 1.0, SC&A determined that the monitoring program appeared to 

target the workers most at risk (technical staff and mechanics), while also covering a crosssection of other occupations that could also be exposed. Table 12 shows the number and 

percentage of workers in each occupation monitored for plutonium intakes, as compared to the 

total number of workers monitored for plutonium, and the years records are available. 

Based on Table 12, it appears that monitoring for plutonium among the different occupations 

mirrors the overall site data presented in Section 1.0, with the Mechanics (38%) and Technical 

Staff (27%) receiving the most plutonium monitoring while other, but not all, occupations were 

monitored. It is noteworthy that no monitoring data are available for Mechanics and Technical 

Staff in the mid-1970s despite plutonium operations occurring at the NMDF, per Table 9. 
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Observation 8: While there are temporal gaps in the job title coverage for mechanics and 

technical staff, less than 1⁄4 of the monitored workers could be identified with a particular 

occupation. Of the monitored workers for which job title information exists, mechanics and 

technical staff constitute 65% of the population. Therefore, it is reasonable to assume that these 

higher-risk job types were not systematically excluded and were likely targeted for plutonium 

monitoring. 

Table 12. Occupations Monitored for Plutonium Intakes at SSFLa

Occupational Group Number of 

Individualsb

Years Monitoring Data 

Available

Percentage of Individuals in 

an Occupational Group

Construction [redacted] 1975 2.1%

Electrician [redacted] 1970–1971, 1975 6.3%

Fireman 0 - 0%

Health and Safety/Health Physicsc[redacted] 1965–1976, 1984–1985 10.4%

Inspector [redacted] 1987–1988, 1990 2.1%

Janitor 0 - 0%

Machinist 0 - 0%

Maintenanced0 - 0%

Management 0 - 0%

Mechanice18 1967–1973, 1978–1990 37.5%

Office Personnel [redacted] 1983, 1986 2.1%

Operator 0 - 0%

Quality Assurance [redacted] 1985 2.1%

Radiographer 0 - 0%

Reactor Operator 0 - 0%

Security 0 - 0%

Shipping 0 - 0%

Technical Stafff 13 1967–1976, 1984 27.1%

Technician [redacted] 1987 2.1%

Truck Driver [redacted] 1983–1984 2.1%

Unknowng[redacted] 1968–1971, 1974–1979, 

1983–1985, 1987, 1989

6.3%

Welder 0 - 0%

Total 48 - 100%

a

 Excludes dose records prior to 1965 and between January 1991 and mid-June 1993.

b

 An individual may be represented more than once if their occupation changed over the course of their 

monitoring.

c

 Includes medical staff, and health physicists and their support staff

d

 Includes individuals with ‘maintenance’ as part of their job title, utility services staff, and painters

e

 Includes individuals with ‘mechanic’ somewhere in their job title, and those who describe themselves as reactor 

and reactor fuel assemblers/disassemblers.

f

 Includes individuals with ‘engineer’ as part of their job title, physicists, chemists, and analysts

g

 Either a job title was not specified and marked as unknown, or it is unclear what the job title represents (for 

example, ‘Y&S’, ‘FMS’).
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4.0 REVIEW OF FISSION PRODUCT DATA AND ASSOCIATED 

COWORKER INTAKES 

Section 4 provides a detailed review of the fission product coworker model presented in 

ORAUT-OTIB-0080. Specifically, the OPOS values are examined for applicability and 

appropriateness in coworker modeling (Section 4.1). The raw data were also used to try and 

recreate the OPOS results presented in Attachment 1 of ORAUT-OTIB-0080 (Section 4.2). 

Section 4.3 presents an analysis of the occupations monitored for internal fission product 

exposures. 

4.1 REVIEW OF CALCULATED OPOS RESULTS 

The calculated OPOS excretion rates for fission products were calculated on an annual basis for 

the purpose of coworker intake modeling from 1965–1991. SC&A was provided with the raw 

database used by ORAUT in calculation of fission product OPOS values shown in Table A-3 of 

ORAUT-OTIB-0080. SC&A extracted the appropriate sample results based on the criteria 

described in Section 1 of this report. It was immediately clear that the number of monitored 

workers was significantly different between the two compilations in the earlier years. A 

comparison of the number of OPOS results per evaluation period is shown in Table 13. As is 

evident from the table, the two compilations of data do not agree until about 1975, at which point 

the two compilations become nearly identical as far as how many monitored workers exist. 

Table 13. Comparison of SC&A and NIOSH OPOS Totals by Year 

Year SC&A OPOS 

Total

NIOSH OPOS 

Total Year SC&A OPOS Total

NIOSH OPOS 

Total

1965 52 321 1979 84 84

1966 24 216 1980 79 82

1967 21 221 1981 55 55

1968 136 154 1982 65 65

1969 111 116 1983 45 45

1970 44 125 1984 65 65

1971 0 84 1985 67 67

1972 0 75 1986 78 78

1973 0 66 1987 63 63

1974 51 55 1988 65 65

1975 83 84 1989 50 50

1976 67 67 1990 33 33

1977 75 75 1991 27 27 1978 118 118

ORAUT-OTIB-0080 states the following concerning which records to use for fission product 

coworker analysis: 

The analysis type designations that were used were ‘MFP’ (mixed fission product, 

chemical separation of alkaline earths and rare earths including strontium) if the 

method type was “B” for beta counting (which excludes gamma-counting data), 

‘MFPB’, ‘MFP(B),’ and ‘FP’ if the method type was ‘3A’ (mixed fission 

products less cesium and volatiles, assumed to indicate strontium). Sufficient 
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bioassay data were available to perform a statistical analysis for 1965 through 

1991. [Emphasis added.] 

In its OPOS calculation, SC&A also considered the following sample types as “fitting” the stated 

criteria for inclusion: 

 Nuclide/Analysis: MFP, Method Type: 7B  1 sample 

 Nuclide/Analysis: MFP, Type 5B  1 sample 

 Nuclide/Analysis: MFP, Type 3B  1 sample 

 Nuclide/Analysis: MFP, Method Type: 2B (this constituted the vast majority of samples 

with a Nuclide/Analysis: MFP) 1170 samples 

 Nuclide/Analysis: MFP, Method Type: 1B 1 sample 

 Nuclide/Analysis: FP, Method Type: 3AG  2 samples 

 Nuclide/Analysis: FP, Method Type: 3A T-U  1 sample 

SC&A noted that additional codes with a “Nuclide/Analysis” type of either FP or MFP were 

found that did not fit the above list of criteria for inclusion. These samples included the 

following “Nuclide/Analysis” type and “Method Type” as shown in Table 14. 

Table 14. Additional Sample Types Not Meeting the Criteria for Inclusion per 

ORAUT-OTIB-0080 

“Nuclide/Analysis” Type “Method” Type Number of Samples

FP 3B 2,737

MFP Blank 2,388

MFP 3A 10

FP Blank 9

MFPG Blank 5

MFP G 3

FP-3AG-Co-60 Blank 1

MFP 1D 1

MFP(G) Blank 1

SC&A added the datapoints in Table 14 to the original datapoints that fit the ORAUT-OTIB0080 criteria and compared the number of workers by year for this expanded dataset. The 

comparison of the number of monitored workers by year for the expanded SC&A dataset versus 

the NIOSH compilation in Table A-3 is shown in Table 15. With the exception of minor 

differences in a handful of years, the expanded SC&A dataset matches the number of monitored 

workers in each year very well. Note that the total number of workers in the earlier years (left 

side of the table) now matches exactly, where this was the biggest discrepancy shown in 

Table 13. Conversely, years that had matched perfectly in Table 13 in the later years (right side 

of the table) are now off by one or two workers. It is likely that not all of the samples tabulated 

in Table 13 as being “additional” were actually used in NIOSH’s calculation. 

Finding 7: Based on the number of monitored workers tabulated in Tables 13 and 14, it is likely 

that NIOSH did not adhere to the inclusion guidelines outlined in ORAUT-OTIB-0080. It 
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appears that all samples labelled “FP” or “MFP” in the “Nuclide/Analysis” column were 

included, regardless of the actual designated method of measurement. 

Table 15. Comparison of Expanded SC&A OPOS Totals and NIOSH OPOS 

Totals by Year 

Year SC&A OPOS 

Total

NIOSH OPOS 

Total Year SC&A OPOS Total

NIOSH OPOS 

Total

1965 321 321 1979 85 84

1966 216 216 1980 82 82

1967 221 221 1981 55 55

1968 154 154 1982 67 65

1969 116 116 1983 45 45

1970 124 125 1984 66 65

1971 84 84 1985 68 67

1972 75 75 1986 79 78

1973 66 66 1987 63 63

1974 55 55 1988 65 65

1975 84 84 1989 51 50

1976 67 67 1990 33 33

1977 75 75 1991 27 27

1978 118 118

4.2 REPLICATION OF CALCULATED OPOS RESULTS 

SC&A attempted to recreate the NIOSH OPOS values contained in ORAUT-OTIB-0080 

Table A-3 based on the expanded dataset described above. The resulting comparison of the 

calculated geometric mean (GM) and 84th percentile is shown in Table 16. It is clear from this 

table that SC&A-calculated OPOS results and NIOSH-calculated values do not match up for 

most years. In the early years, the NIOSH OPOS result is generally higher, while in the latter 

years the SC&A result is higher. Sometimes the difference between the calculated 50th

percentile by year was a factor of 3–6.

Finding 8: SC&A was not able to duplicate the FP OPOS values presented in Table A-3 of 

ORAUT-OTIB-0080 to within a reasonable level of precision for most years. The source of the 

discrepancies should be identified and resolved prior to the calculation of coworker intakes. 

Table 16. Comparison of SC&A-Calculated OPOS Values Using Expanded Fission 

Product Data to NIOSH OPOS Values contained in Table A-3 of ORAUT-OTIB-0080

Year

Effective 

Bioassay Date

SC&A 50th

Percentile OPOS 

(dpm/1.4L)

NIOSH 50th

Percentile OPOS 

(dpm/1.4L)

SC&A 84th

Percentile OPOS 

(dpm/1.4L)

NIOSH 84th

Percentile OPOS 

(dpm/1.4L)

1965 7/1/1965 15.82 26.06 51.63 58.88

1966 7/1/1966 18.94 21.04 38.84 38.26

1967 7/1/1967 19.64 20.99 44.30 44.65

1968 7/1/1968 5.26 19.63 14.35 30.48

1969 7/1/1969 1.49 10.07 9.25 24.71

1970 7/1/1970 18.44 19.34 29.40 24.40

1971 7/1/1971 29.13 16.54 32.37 28.26

1972 7/1/1972 28.06 14.67 30.04 23.58
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Table 16. Comparison of SC&A-Calculated OPOS Values Using Expanded Fission 

Product Data to NIOSH OPOS Values contained in Table A-3 of ORAUT-OTIB-0080

Year

Effective 

Bioassay Date

SC&A 50th

Percentile OPOS 

(dpm/1.4L)

NIOSH 50th

Percentile OPOS 

(dpm/1.4L)

SC&A 84th

Percentile OPOS 

(dpm/1.4L)

NIOSH 84th

Percentile OPOS 

(dpm/1.4L)

1973 7/1/1973 27.79 12.90 31.76 24.28

1974 7/1/1974 20.23 4.65 37.65 22.27

1975 7/1/1975 21.20 12.90 33.39 20.51

1976 7/1/1976 17.98 18.72 26.77 29.14

1977 7/1/1977 21.24 6.98 43.84 31.58

1978 7/1/1978 20.59 14.25 42.32 30.33

1979 7/1/1979 20.10 9.06 39.43 22.97

1980 7/1/1980 15.30 5.54 27.34 17.07

1981 7/1/1981 17.24 17.75 27.74 33.77

1982 7/1/1982 24.65 14.67 44.33 33.51

1983 7/1/1983 20.66 15.42 36.01 23.90

1984 7/1/1984 24.17 3.59 44.38 12.29

1985 7/1/1985 26.37 15.89 42.08 26.69

1986 7/1/1986 27.66 16.15 41.77 27.12

1987 7/1/1987 26.16 21.92 43.38 38.70

1988 7/1/1988 33.90 24.52 43.25 29.45

1989 7/1/1989 32.86 22.79 45.78 37.83

1990 7/1/1990 23.47 7.67 47.77 45.77

1991 3/1/1991 28.06 8.08 47.24 35.30

Observation 9: Note that the last effective bioassay date in Table 15 is March 1, 1991. 

Examination of the available data indicates that FP bioassay data effectively end in mid-April. 

This comports with an effective bioassay date for the evaluation period of January 1, 1991, to 

roughly April 30, 1991. Since the end of April 1991 marks the end of the FP bioassay 

evaluation, the calculated intakes found in ORAUT-OTIB-0080 Table 5-6 should also end in 

April 1991, not December 1991. However, it is not clear if changing the date of the end of the 

chronic intake calculation would significantly affect the calculated intakes. 

NIOSH has assumed a single chronic intake regime over all 27 years of exposure. SC&A 

acknowledges that the choice of intake regimes involves some professional judgment on the part 

of the dose modeler. However, if the intake regimes had been split up into smaller parts, the 

calculated intake rates could be significantly higher (such as during the periods of 1965–1968 or 

1985–1989). In general, the longer the period of time one assumes for the chronic intake regime, 

the lower the overall calculated intakes will be. The justification for grouping certain years 

together into a single intake regime is often that the magnitude of the bioassay results are 

sufficiently similar that the exposure potential during that time is also similar. 

Figure 8 plots the 50th percentile NIOSH OPOS value by year; as can be seen, the range in 

magnitude of the 50th percentile bioassay value can vary significantly. For example, the 

difference in the calculated result in 1984 (3.59 dpm/1.4L) and surrounding years is 

approximately a factor of 4. The difference of a factor of 4 is not dissimilar to the plutonium 

excretion differences in the first and second intake regimes in Table 6 (0.064/0.026 = factor of 

2.5), which warranted a change in intake regime. SC&A admits that this is not an issue that is 
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restricted to the SSFL dataset, but has programmatic implications as well. It would be beneficial 

to provide the rationale in the context of historical site operations, exposure potential, and 

observed OPOS variability to lend scientific credibility to the chosen approach of a single intake 

regime for mixed fission products. 

Figure 8. Calculated NIOSH OPOS Values by Year 

Finding 9: The coworker model would benefit from a substantive discussion providing the 

rationale for combining all 27 years into a single intake regime. This might include a discussion 

on the variability in the calculated measurements observed by year, changes in the GSD 

indicating a larger spread of potential exposures, the actual effect on the calculated intakes in 

relation to dose, and changes in site operations. Such a discussion would reinforce the 

assumption that a single chronic intake regime is appropriate over a significantly long period of 

time. 

4.3 OCCUPATIONS MONITORED FOR FISSION PRODUCT INTAKE 

Occupational information was analyzed by cross-referencing dose records that specify an 

analytical method for detecting mixed fission products with occupational information developed 

as described in Section 2.0. From Section 2.0, SC&A determined that the monitoring program 

appeared to target the workers most at risk (technical staff and mechanics), while also covering a 

cross-section of other occupations that could also be exposed. Table 17 shows the number and 

percentage of workers in each occupation monitored for mixed fission product intakes as 

compared to the total number of workers monitored for mixed fission products.
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Table 17. Occupations Monitored for Mixed Fission Product Intakes at SSFLa

Occupational Group Number of 

Individualsb Years Monitoring Data Available Percentage of Individuals in an Occupational Group

Construction [redacted] 1975 <1%

Electrician [redacted] 1974–1976, 1980 3%

Fireman [redacted] 1968–1970, 1974 3%

Health and Safety/

Health Physicsc

[redacted] 1968–1970, 1974–1977, 1979–1980, 

1984–1985, 1988

4%

Inspector [redacted] 1988–1990 <1%

Janitor [redacted] 1968–1970 4%

Machinist [redacted] 1965–1966 <1%

Maintenanced[redacted] 1968–1969 3%

Management [redacted] 1965–1969, 1977–1981, 1986 3%

Mechanice41 1965–1970, 1974–1991 34%

Office Personnel [redacted] 1968, 1970, 1975, 1979–1982, 1985–

1989

2%

Operator [redacted] 1968–1969 2%

Quality Assurance [redacted] 1985–1986 <1%

Radiographer 0 - 0%

Reactor Operator [redacted] 1965–1969, 1981, 1983 3%

Security [redacted] 1968–1970, 1986, 1989 4%

Shipping [redacted] 1968 <1%

Technical Stafff 32 1965–1970, 1974–1991 27%

Technician [redacted] 1978–1981, 1984–1989 3%

Truck Driver [redacted] 1975–1979, 1981–1986 <1%

Unknowng[redacted] 1965, 1977–1991 3%

Welder [redacted] 1977–1978, 1982 2%

Total 120 - 100%

a

 Excludes dose records prior to 1965 and between January 1991 and mid-June 1993.

b

 An individual may be represented more than once if their occupation changed over the course of their monitoring.

c

 Includes medical staff, and health physicists and their support staff

d

 Includes individuals with ‘maintenance’ as part of their job title, utility services staff, and painters

e

 Includes individuals with ‘mechanic’ somewhere in their job title, and those who describe themselves as reactor and 

reactor fuel assemblers/disassemblers.

f

 Includes individuals with ‘engineer’ as part of their job title, physicists, chemists, and analysts

g

 Either a job title was not specified and marked as unknown, or it is unclear what the job title represents (for 

example, ‘Y&S’, ‘FMS’).

Based on Table 17, it appears that monitoring for mixed fission products among the different 

occupations mirrors the overall site data presented in Section 1.0, with the Mechanics (38%) and 

Technical Staff (27%) receiving the most plutonium monitoring while other, but not all, 

occupations were monitored. The distribution of monitoring among the occupations also closely 

resembles the distribution of occupations monitored for plutonium when compared to Table 12. 

The annual distribution of monitoring among Mechanics and Technical Staff appears to 

correspond to NMDF operations, which would explain the lack of monitoring data between 1970 

and 1974. 
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5.0 REVIEW OF URANIUM DATA AND ASSOCIATED 

COWORKER INTAKES

Section 5 provides a detailed review of the uranium coworker model presented in ORAUTOTIB-0080. Specifically, the OPOS values are examined for applicability and appropriateness 

in coworker modeling (Section 5.1). The raw data were also used to try and recreate the OPOS 

results presented in Attachment 1 of ORAUT-OTIB-0080. Section 5.2 presents an analysis of 

the assumptions associated with uranium uptake, and whether SSFL bioassay collection and 

analysis practices were fully considered. Section 5.3 presents an analysis of the occupations 

monitored for internal uranium exposures.

5.1 REVIEW AND REPRODUCTION OF CALCULATED OPOS RESULTS 

The OPOS excretion rates for uranium bioassay were calculated on an annual basis for the 

purpose of coworker intake modeling from 1965–1988. SC&A was provided with the raw 

database used by ORAUT in calculation of uranium OPOS values shown in Table A-2 of 

ORAUT-OTIB-0080. SC&A extracted the appropriate sample results based on the criteria 

described in Section 1 of this report. SC&A attempted to replicate the calculated OPOS values 

in Table A-2 of ORAUT-OTIB-0080; the results are contained in Table 18. 

It is evident from the table that the number of monitored workers contained in both SC&A’s and 

NIOSH’s OPOS analysis are identical in all years but two (1967 and 1987). The actual OPOS 

results between the two calculations are rather close in most years; however, some years display 

over an order of magnitude difference (1968, 1969, 1975 and 1982). It is not clear from the 

available data where the discrepancies arise. 

Table 18. Comparison of SC&A-Calculated OPOS Results to NIOSH-Calculated 

OPOS Results Contained in ORAUT-OTIB-0080 Table A-2

Year

Effective 

Bioassay 

Date

SC&A 

OPOS

NIOSH 

OPOS

SC&A 50th

Percentile 

OPOS 

(dpm/1.4L)

NIOSH 50th

Percentile 

OPOS 

(dpm/1.4L)

SC&A 84th

Percentile 

OPOS 

(dpm/1.4L)

NIOSH 84th

Percentile 

OPOS 

(dpm/1.4L)

1965 7/1/1965 470 470 2.22 2.90 7.77 8.33

1966 7/1/1966 298 298 4.74 4.52 11.97 10.83

1967 7/1/1967 336 338 5.65 5.47 15.39 14.69

1968 7/1/1968 269 269 0.70 6.22 6.06 14.05

1969 7/1/1969 153 153 0.35 2.50 3.19 8.6

1970 7/1/1970 164 164 6.63 3.41 10.42 7.79

1971 7/1/1971 135 135 8.50 2.35 8.80 5.09

1972 7/1/1972 122 122 8.83 1.68 10.05 5.67

1973 7/1/1973 88 88 6.67 4.99 9.19 7.09

1974 7/1/1974 86 86 4.96 5.95 7.74 7.53

1975 7/1/1975 134 135 4.37 0.23 7.44 1.13

1976 7/1/1976 166 166 6.04 5.77 8.46 8.03

1977 7/1/1977 95 95 5.89 4.52 8.72 6.68

1978 7/1/1978 147 147 4.16 4.89 7.39 14.14

1979 7/1/1979 124 124 4.96 5.14 7.49 6.52
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Table 18. Comparison of SC&A-Calculated OPOS Results to NIOSH-Calculated 

OPOS Results Contained in ORAUT-OTIB-0080 Table A-2

Year

Effective 

Bioassay 

Date

SC&A 

OPOS

NIOSH 

OPOS

SC&A 50th

Percentile 

OPOS 

(dpm/1.4L)

NIOSH 50th

Percentile 

OPOS 

(dpm/1.4L)

SC&A 84th

Percentile 

OPOS 

(dpm/1.4L)

NIOSH 84th

Percentile 

OPOS 

(dpm/1.4L)

1980 7/1/1980 109 109 2.91 1.76 6.00 3.66

1981 7/1/1981 104 104 2.61 1.34 3.44 2.08

1982 7/1/1982 74 75 2.27 0.22 3.40 0.86

1983 7/1/1983 57 57 2.43 3.17 3.44 3.41

1984* 7/1/1984 12 Not Listed 2.52 Not Listed 4.15 Not Listed

1985* 7/1/1985 2 Not Listed Insufficient 

Data Not Listed Insufficient Data Not Listed

1986 7/1/1986 34 34 2.90 2.05 3.58 3.18

1987 7/1/1987 58 59 2.52 1.48 3.54 2.29

1988 7/1/1988 63 63 2.46 1.44 3.48 2.24

*These years were not included in the ORAUT-OTIB-0080 coworker analysis 

Finding 10: SC&A’s attempt to duplicate the uranium OPOS values presented in Table A-2 of 

ORAUT-OTIB-0080 was off by an order of magnitude in 4 of the 25 years analyzed by NIOSH 

(1984 and 1985 were not analyzed). The source of the discrepancies should be identified and 

resolved prior to the final calculation of coworker intakes. 

As evidenced in Table 18 and also from Section 4.4 of ORAUT-OTIB-0080, NIOSH did not 

include the years 1984 and 1985 in the coworker analysis. Additionally, it was determined that 

1975 results were not contemporaneous with the surrounding years and so were excluded. 

Specifically, ORAUT-OTIB-0080 states: 

In 1974,7the excretion rates from the statistical analysis were inconsistent with 

the contemporaneous years and were excluded to be favorable to claimants. In 

1984 and 1985, the results are zero (statistical analysis not possible) and were 

similarly excluded. 

7

 This appears to be a typo and should be “1975” 

It is assumed that the above statement meant to indicate 1975 as not being contemporaneous with 

surrounding years (1974 - 5.95 dpm/1.4L; 1975 - 0.23 dpm/1.4L; 1976 - 5.77 dpm/L). It should 

be noted that SC&A calculated an OPOS value of 4.37 dpm/1.4L for this year, which is 

contemporaneous with the surrounding years. 

 

With regard to 1984 and 1985, SC&A agrees that it is likely not statistically appropriate to 

analyze the 1984 and 1985 data for the purposes of a coworker model because of the paucity of 

available samples (12 and 2 OPOS results per year). The 1984 and 1985 data are tabulated in 

Table 19 for reference. Only one worker was sampled twice in that 2-year period. 

 
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Table 19. Tabulation of Available 1984 and 1985 Data 

Date of Sample Sample Result

(dpm/1.4L) Date of Sample Sample Result(dpm/1.4L)

1/1/1984 < 3.547 7/1/1984 < 3.267

2/13/1984 < 3.547 8/8/1984 < 0.215

2/14/1984 < 3.547 11/19/1984 < 3.547

2/14/1984 < 3.547 11/28/1984 < 3.547

2/14/1984 < 3.547 11/28/1984 < 3.547

2/21/1984 0 2/11/1985 0

4/1/1984 < 3.547 4/18/1985 < 3.5

4/9/1984 < 3.547

 

Based on the data shown in Table 19, there does not appear to be any discernible reason why 

samples were taken on the dates shown (such as in reaction to some unforeseen exposure or 

perhaps a trip to another site). The more important question might be, why there are so few 

uranium urinalysis data for these 2 years? SC&A was not able to identify any obvious rationale 

for why there are so few samples taken in 1984 and 1985. A chart of the number of samples and 

monitored workers for plutonium, uranium, and fission products is shown in Figures 9 and 10.

Figure 9. Comparison of the Number of Samples for Coworker Radionuclides 

during the 1980s 
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Figure 10. Comparison of the Number of Monitored Workers for Coworker Radionuclides 

during the 1980s 

During the 1980s, plutonium sampling was at its highest from 1983–1985. Fission product 

sampling increased incrementally from 1983 to 1985. Based on Figure 10, it appears that the 

number of monitored plutonium and fission product workers was also increased during 1984 and 

1985. The trend of little-to-no monitoring results for uranium was not mirrored by the 

monitoring for other contaminants of concern. 

Finding 11: It is important to understand the reasons for the drastic decrease in observed 

sampling for uranium from 1984 to 1985. Concurrently, because the paucity of the data makes it 

impossible to make quantitative statements about the available bioassay data from 1984 to 1985, 

it should be established that conditions and exposure potential are sufficiently similar to (or 

bounded by) surrounding years to validate the use as surrogate data. 

5.2 REVIEW OF URANIUM MONITORING PRACTICES AND INCORPORATION 

INTO INTERNAL MODEL 

SC&A examined how uranium exposures were calculated by NIOSH in relation to SSFL 

bioassay sample collection and analysis practices. 

Based on information in ORAUT-TKBS-0038-5, Rev. 01 (ORAUT 2010, pg. 14), routine urine 

samples were collected after 48 hours’ absence of work (it is unclear whether employees were 

requested to fill the samples on Sunday during the entire SSFL operational history): 
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By 1959, routine urine samples were requested on Fridays, and each monitored

employee was required to submit the first voiding on Monday morning (after an

absence from work of 48 hr or more). The time of the previous voiding was 

recorded to determine the excretion rate. If the Monday morning sample was 

verified as positive, a series of 24-hr samples was collected to determine the body 

burden. Employees were requested to fill these samples on Sunday. Appropriate 

adjustments to this schedule were made for weekend work, etc.

Table 20 provides a description of the sampling protocol at Santa Susana (from ORAUT 2010, 

Table 5-7.) 

Table 20. Description of Bioassay Sampling Protocol at SSFL 

Media Sample

Urine Routine samples. Single voiding collected on Mdeosndcriaptyiomnorning before returning to work. Collected in 

single bottle. Also referred to as rate samples. Positive samples were followed up with additional rate

sample for verification. Time of sample collection and previous voiding were recorded. Normally 

collected in a 16-oz container. By 1999, 1-L polyethylene containers were used. It appears that one 

full container (900–1,000 mL) was collected and collection time noted.

Urine 24-hr samples. Used for follow-up to verify rate samples or for incidents. One or more samples could

be requested. Single samples were collected at home on Sunday. Collected in 32-oz or 1-L 

polyethylene containers.

Urine Spot samples. For follow-up to incidents, spot sample could be collected as soon as possible. This 

sample was probably collected in 16-oz container.

Feces Could be requested in conjunction with urine samples as follow-up to incidents. No descriptions of 

historical fecal kits were located. By 1999, single voiding samples were collected in 83-oz 

polyethylene containers. The minimum mass that was considered adequate was 30 g.

Source: ORAUT 2010, Table 5-7 

SC&A examined the available data to determine how sample collections were distributed during 

individual weeks; the results are shown in Table 21. Since the “Date Received” column in the 

dose records represents when the sample was received by the lab, only samples that had a date 

listed in the “Sample Date (collection)” column were analyzed in Table 21. Bioassay results 

with “Date Received” are analyzed in Table 22. 

Table 21. Distribution of Bioassay Results by the Day of the Week Sample is Collected 

Day of the Week Uranium Plutonium Fission Products

Sunday 489 (9.23%) 345 (39.75%) 424 (12.68%)

Monday 3,029 (57.18%) 227 (26.15%) 1,566 (46.82%)

Tuesday 765 (14.44%) 70 (8.06%) 482 (14.41%)

Wednesday 359 (6.78%) 75 (8.64%) 276 (8.25%)

Thursday 271 (5.12%) 65 (7.49%) 244 (7.29%)

Friday 321 (6.06%) 57 (6.57%) 284 (8.49%)

Saturday 63 (1.19%) 63 (7.26%) 69 (2.06%)

Total 5,297 (100%) 868 (100%) 3,345 (100%)

As shown in Table 21, between 60% and 66% of the available bioassay data that provided a 

sample collection date was taken on either Sunday or Monday. This reinforces the notion that 

the majority of samples were taken after approximately a 48-hour absence from work. 
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For completeness, Table 22 presents the distribution of samples (by day) for bioassay entries that 

only provide a “Received Date.” As seen in the table, the majority of samples for uranium and 

plutonium were received by the lab on Wednesday or Thursday (~80% for uranium and ~76% 

for plutonium). This trend was not observed for fission products, where the majority of samples 

were “received” on Monday. ORAUT-OTIB-0080 states the following concerning the “sample 

received” date: 

The sample receipt date on the laboratory data form is usually a few days after 

the sample collection date on the facility form.

This is consistent with a bioassay protocol in which the majority of samples were collected on 

Monday and received a few days later at the analytical laboratory (Wednesday or Thursday). 

Table 22. Distribution of Bioassay Results by “Sample Received” Date

Day of the Week Uranium Plutonium Fission Products

Sunday 11 (0.33%) 3 (8.82%) 5 (15.15%)

Monday 152 (4.51%) 3 (8.82%) 18 (54.55%)

Tuesday 145 (4.30%) 0 (0.00%) 2 (6.06%)

Wednesday 1,179 (34.94%) 11 (32.35%) 2 (6.06%)

Thursday 1,521 (45.08%) 15 (44.12%) 2 (6.06%)

Friday 363 (10.76%) 1 (2.94%) 3 (9.09%)

Saturday 3 (0.09%) 1 (2.94%) 1 (3.03%)

Total 3,374 (100%) 34 (100%) 33 (100%)

In ORAUT-OTIB-0080, Rev. 00, NIOSH assumes a continuous intake pattern to calculate 

uranium coworker intake rates. If it is assumed that all urine samples were collected after a 

48-hour absence from work, the approach used by NIOSH is underestimating the intake rates. 

For Type F uranium, NIOSH underestimates the intake rate by a factor of around 4. For Type M 

uranium, the underestimation is a factor of around 2, and for Type S the underestimation factor 

fluctuates from 1.6 to 2.0. A description of SC&A’s analysis for each type of uranium follows.

For Type F uranium, NIOSH calculated GM intake rates to be (ORAUT-OTIB-0080, Table 5-3):

1965–1968 17.6 dpm/d 

1969–1972 8.91 dpm/d

1973–1979 18.87 dpm/d 

1980–1988 5.666 dpm/d 

Figure 11 shows the predicted uranium bioassay results calculated using these derived GM

intakes and their comparison with the 50th percentile bioassay results presented in ORAUTOTIB-0080. 
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Source: ORAUT-OTIB-0080, Figure A-13

Figure 11. Predicted Uranium Bioassay Results Calculated using IMBA-derived Uranium 

Intake Rates (Line) compared with Bioassay Results (Dots), 50th Percentile, 

All Years, Type F 

In Table 23 (2nd and 3rd columns), SC&A reproduced the values of Figure 11 using the 

assumption of a continuous intake. In addition, Table 23 shows the predicted uranium bioassay 

values (column 4) for samples taken on Monday mornings (after a 48-hour absence from work). 

As shown in column 5, the predicted excretion rates on Monday mornings are a factor of about 4 

smaller than the ones predicted using a continuous intake rate. This means that the intake rates 

calculated for Type F uranium are underestimated by a factor of about 4. 

Tables 24 and 25 are equivalent to Table 23, using uranium compounds of Type M and Type S, 

respectively. 

Table 23. Predicted Urine Excretion Rates Corresponding to NIOSH’s Derived GM

Intake Rates of Type F Uranium Compounds, Assuming a Continuous Exposure as 

compared to a 5-day Per Week Exposure, with Monday Morning Sample Collections

Year

(OTIB-0080 Intake rate)

50th percentile

urine (dpm/d)

Predicted 24-Hour

Urine (dpm) Assuming 

a Continuous Intake

Predicted 24-Hour Urine 

(dpm/d) Assuming Monday 

Morning Collection (dpm/d)

Underestimation 

of Intake*

1965 (17.6 dpm/d) 2.90 4.78 1.10 4.35

1966 (17.6 dpm/d) 4.52 4.80 1.11 4.32

1967 (17.6 dpm/d) 5.47 4.81 1.13 4.26

1968 (17.6 dpm/d) 6.22 4.83 1.13 4.27

1969 (8.91 dpm/d) 2.50 2.42 0.56 4.35

1970 (8.91 dpm/d) 3.41 2.43 0.56 4.31

1971 (8.91 dpm/d) 2.35 2.44 0.57 4.28

1972 (8.91 dpm/d) 1.68 2.44 0.57 4.25

1973 (18.87 dpm/d) 4.99 5.12 1.18 4.34

1974 (18.87 dpm/d) 5.95 5.15 1.19 4.33

1975 (18.87 dpm/d) 0.23 5.16 1.21 4.26

1976 (18.87 dpm/d) 5.77 5.18 1.22 4.25

1977 (18.87 dpm/d) 4.52 5.19 1.23 4.22

1978 (18.87 dpm/d) 4.89 5.20 1.23 4.23

1979 (18.87 dpm/d) 5.14 5.21 1.24 4.20

1980 (5.666 dpm/d) 1.76 1.52 0.35 4.29
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Table 23. Predicted Urine Excretion Rates Corresponding to NIOSH’s Derived GM

Intake Rates of Type F Uranium Compounds, Assuming a Continuous Exposure as 

compared to a 5-day Per Week Exposure, with Monday Morning Sample Collections

Year

(OTIB-0080 Intake rate)

50th percentile

urine (dpm/d)

Predicted 24-Hour

Urine (dpm) Assuming 

a Continuous Intake

Predicted 24-Hour Urine 

(dpm/d) Assuming Monday 

Morning Collection (dpm/d)

Underestimation 

of Intake*

1981 (5.666 dpm/d) 1.34 1.53 0.36 4.26

1982 (5.666 dpm/d) 0.22 1.53 0.36 4.23

1983 (5.666 dpm/d) 3.17 1.54 0.37 4.22

1984 (5.666 dpm/d) N/A 1.54 0.37 4.18

1985 (5.666 dpm/d) N/A 1.54 0.37 4.16

1986 (5.666 dpm/d) 2.05 1.55 0.37 4.17

1987 (5.666 dpm/d) 1.48 1.55 0.37 4.14

1988 (5.666 dpm/d) 1.44 1.55 0.38 4.12

* NIOSH-derived geometric mean intake rates should be multiplied by the values under the column 

“Underestimation of Intakes” in order to reproduce NIOSH’s predicted bioassay results which were derived by 

NIOSH as the best fit to the 50th percentile bioassay results, given in Table A-2 and Figure A-13 of OTIB-0080 

Rev. 0 (ORAUT 2014a).

Table 24. Predicted Urine Excretion Rates Corresponding to NIOSH’s Derived GM Intake 

Rates of Type M Uranium Compounds, Assuming a Continuous Exposure as compared to 

a 5-day Per Week Exposure, with Monday Morning Sample Collections 

Year

(OTIB-0080 Intake rate)

50th percentile

urine (dpm/d)

Predicted 24-Hour

Urine (dpm) Assuming 

a Continuous Intake

Predicted 24-Hour Urine 

(dpm/d) Assuming Monday 

Morning Collection (dpm/d)

Underestimation 

of Intake*

1965 (17.6 dpm/d) 2.90 4.77 2.14 2.23

1966 (17.6 dpm/d) 4.52 5.03 2.33 2.16

1967 (17.6 dpm/d) 5.47 5.07 2.36 2.15

1968 (17.6 dpm/d) 6.22 5.09 2.37 2.15

1969 (8.91 dpm/d) 2.50 2.19 0.98 2.23

1970 (8.91 dpm/d) 3.41 2.31 1.07 2.16

1971 (8.91 dpm/d) 2.35 2.33 1.08 2.16

1972 (8.91 dpm/d) 1.68 2.34 1.09 2.15

1973 (18.87 dpm/d) 4.99 4.90 2.20 2.23

1974 (18.87 dpm/d) 5.95 5.17 2.39 2.16

1975 (18.87 dpm/d) 0.23 5.22 2.42 2.16

1976 (18.87 dpm/d) 5.77 5.23 2.44 2.14

1977 (18.87 dpm/d) 4.52 5.25 2.45 2.14

1978 (18.87 dpm/d) 4.89 5.26 2.45 2.14

1979 (18.87 dpm/d) 5.14 5.27 2.46 2.14

1980 (5.666 dpm/d) 1.76 1.38 0.62 2.23

1981 (5.666 dpm/d) 1.34 1.45 0.67 2.16

1982 (5.666 dpm/d) 0.22 1.47 0.68 2.16

1983 (5.666 dpm/d) 3.17 1.47 0.69 2.15

1984 (5.666 dpm/d) N/A 1.47 0.69 2.14

1985 (5.666 dpm/d) N/A 1.48 0.69 2.14

1986 (5.666 dpm/d) 2.05 1.48 0.69 2.14

1987 (5.666 dpm/d) 1.48 1.48 0.69 2.14

1988 (5.666 dpm/d) 1.44 1.48 0.70 2.13

* NIOSH-derived geometric mean intake rates should be multiplied by the values under the column 

“Underestimation of Intakes” in order to reproduce NIOSH’s predicted bioassay results which were derived by 

NIOSH as the best fit to the 50th percentile bioassay results, given in Table A-2 and Figure A-13 (ORAUT 2014a).
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Table 25. Predicted Urine Excretion Rates Corresponding to NIOSH’s Derived Geometric 

Mean Intake Rates of Type S Uranium Compounds, Assuming a Continuous Exposure as 

compared to a 5-day Per Week Exposure, with Monday Morning Sample Collections 

Year

(OTIB-0080 Intake rate)

50th percentile

urine (dpm/d)

Predicted 24-Hour

Urine (dpm) Assuming 

a Continuous Intake

Predicted 24-Hour Urine 

(dpm/d) Assuming Monday 

Morning Collection (dpm/d)

Underestimation of 

Intake*

1965 (17.6 dpm/d) 2.90 2.83 1.21 2.34

1966 (17.6 dpm/d) 4.52 4.38 2.32 1.89

1967 (17.6 dpm/d) 5.47 5.50 3.11 1.77

1968 (17.6 dpm/d) 6.22 6.32 3.70 1.71

1969 (8.91 dpm/d) 2.50 1.33 0.567 2.35

1970 (8.91 dpm/d) 3.41 2.05 1.08 1.90

1971 (8.91 dpm/d) 2.35 2.57 1.46 1.76

1972 (8.91 dpm/d) 1.68 2.96 1.73 1.71

1973 (18.87 dpm/d) 4.99 2.25 0.97 2.33

1974 (18.87 dpm/d) 5.95 3.49 1.84 1.90

1975 (18.87 dpm/d) 0.23 4.37 2.48 1.76

1976 (18.87 dpm/d) 5.77 5.03 2.94 1.71

1977 (18.87 dpm/d) 4.52 5.52 3.30 1.67

1978 (18.87 dpm/d) 4.89 5.91 3.58 1.65

1979 (18.87 dpm/d) 5.14 6.23 3.80 1.64

1980 (5.666 dpm/d) 1.76 0.69 0.30 2.33

1981 (5.666 dpm/d) 1.34 1.07 0.57 1.89

1982 (5.666 dpm/d) 0.22 1.34 0.76 1.76

1983 (5.666 dpm/d) 3.17 1.54 0.91 1.70

1984 (5.666 dpm/d) N/A 1.70 1.01 1.68

1985 (5.666 dpm/d) N/A 1.82 1.10 1.65

1986 (5.666 dpm/d) 2.05 1.91 1.17 1.63

1987 (5.666 dpm/d) 1.48 1.99 1.22 1.63

1988 (5.666 dpm/d) 1.44 2.06 1.27 1.62

* NIOSH-derived geometric mean intake rates should be multiplied by the values under the column “Underestimation 

of Intakes” in order to reproduce NIOSH’s predicted bioassay results which were derived by NIOSH as the best fit to 

the 50th percentile bioassay results, given in Table A-2 and Figure A-13 of OTIB-0080 Rev. 0 (ORAUT 2014a).

It is worthwhile to note that NIOSH uses a continuous intake to derive coworker intake rates. 

The continuous intake assumes 365 days chronic exposure without rest. In reality, the worker 

has a schedule that consists of periodic breaks that would be characteristic of a particular facility. 

The most common work schedule is a 5-day workweek followed by a 2-day break (weekend). 

The assumption of a chronic nonstop 365-day work period overestimates the worker’s committed 

doses by a factor of 1.4, if in reality the worker has a 5-day workweek (7 days versus 5 days per 

week). In the case of Santa Susana, the committed doses calculated from the radioactivity in 

urine samples taken on Monday mornings after a 48-hour absence from work underestimates the 

dose for Type F and Type M uranium, even if doses were calculated using the 5-days per week 

schedule. It is claimant favorable to assume a continuous intake of 365 work days per year, as is 

assumed for all coworker-derived intakes and assign that value to the unmonitored worker. 

Thus, the uranium intake rates derived by NIOSH should be multiplied by an adjustment factor 

to account for the SSFL protocol of Monday morning sampling. 

The higher underestimate is due to Type F compounds. SC&A calculated 10, 15, 20, 25, 30, 40, 

and 50 years committed equivalent doses for all uranium types (F, M, S), assuming workers were 
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exposed continuously to uranium during the periods 1965–1968, 1969–1972, 1973–1979, and 

1980–1988, with each period of time being analyzed independently. For each period of time, the 

intake rates were the ones defined as GMs by NIOSH in Tables A-6, A-7 and A-8 of OTIB0080. Type S is predominantly the most claimant-favorable type to be chosen, if the correction 

factors for Monday morning samples are not applied. 

Finding 12: Uranium intake rates appear to be underestimated, based on the practice of 

collecting urinalysis samples on Monday morning after 2 days of no exposure. Although not all 

the available bioassay data was collected on Monday morning, NIOSH should apply an 

adjustment factor to the calculated intakes as a claimant-favorable and bounding assumption. 

SC&A would also like to note that it is typically assumed that, for exempt employees, there are 

2,000 work hours per year when determining final intakes; however, there are situations in which 

overtime may be taken into account. In the procedure for determining ambient dose, ORAUTPROC-0060, Rev. 01 (ORAUT 2006b), the Maximizing Methodology stipulates that 2,600 hours 

per year be used, and the Best Estimate Methodology stipulates that 2,500 hours be used. The 

same exposure time should be used across all models when determining dose unless justification 

is provided. 

Observation 10: It is important that assumptions that cross-cut among models (e.g., internal, 

external, ambient) are consistent. Unless there is justification to do otherwise, the length of time 

of exposure of coworkers should be assumed to be the same across all models. Because use of 

overtime is often a common practice among operations personnel, a 2,000 work-year may not 

bound the dose to a worker.

5.3 EXPOSURES TO URANIUM ALUMINIDE 

Attachment A of the Occupational Internal Dose TBD for SSFL (ORAUT 2010) states the 

following concerning uranium aluminide (U-Alx):

SSFL began fabricating reactor fuel elements in the fall of 1965. SSFL fabricated 

fuel for the ATR and the ETR from 1966 to 1968 at the De Soto Facility. In Room 

1110-62 of Building 101 (known at the time as Building 001), briquettes of an 

alloy of 93% EU and aluminum known as uranium aluminide (UAlx) were formed 

in an electric arc melting furnace. These briquettes were crushed to form a 

powder, which was cold-pressed into compacts that became the cores of the 

fabricated fuel plates. The room where these activities took place was known as 

the powder room. 

In 1967 after 15 mo of operation, urine bioassay data indicated that the material 

was probably insoluble and, therefore, that the air activity was not being 

compared to the appropriate MPC. Although the uranium was more than 93% 

235U by weight, 234U accounted for more than 96% of its activity. The insoluble 

MPC for 234U was a factor of 6 lower than the soluble MPC. This led to the 

conclusions that the regulatory standard (weekly average MPC) had been 

exceeded on a number of occasions and that equipment and procedures for 

controlling the airborne uranium were insufficient. 
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An internal investigation determined that the primary reason for the ineffective 

confinement of the uranium was leakage from the crushing glovebox seal, from 

the fume hood of weighing station 2, and from the arc furnace. Temporary 

measures were put in place until more permanent controls could be implemented. 

These changes were put in place and the project was completed. Along with the 

engineering changes, workers were required to wear full-face respirators and 

lapel air samplers (Saxe 1967). At some point, operations were relocated to the 

new powder room. 

... 

A particle size study consisting of two general area air samples from the powder 

room indicated that the particles were less than 1-μm count median diameter 

(CMD) (Alexander 1967). The actual CMD and geometric standard deviation 

were not provided in the reference; therefore, the AMAD could not be determined. 

In the only other particle size reference found, Baurmash (1967) measured the 

particle size distribution for a UO2 grinding operation in the “processing room in 

Bldg. 1.” The CMD reported for this operation was 0.195 μm with a geometric 

standard deviation of 1.66. The mass median diameter was calculated as 

0.42 μm. Using a density of 10.97 g/cm3for UO2, the AMAD for this aerosol is 

1.39 μm. While this information is extremely limited, it points out that small 

particle sizes were at least possible during uranium fuel fabrication. Therefore, 

dose reconstructors should consider a 1-μm AMAD particle size for these 

operations in addition to the default assumption. 

In Attachment H of ORAUT 2010, NIOSH adopts the proposal of Leggett, Eckerman and Boice 

(Leggett et al. 2005), in which mechanical clearance parameters and dissolution parameters of 

the ICRP 66 Human Respiratory Tract Model (ICRP 1994) are modified to fit the bioassay data 

collected from workers who were exposed to U-Alx in the powder room. This proposal states 

that more than 60 persons worked in the powder room from 1966 to the early 1970s, when 

operations were moved to an improved facility. 

SC&A was able to identify 38 workers that were exposed to U-Alx in the powder room during 

1966 and 1967 and who might have worked in exposure conditions above the permissible levels 

of the time. Many of those workers were restricted from work in areas subject to potential 

airborne radioactive material for a certain period of time (Atomics International 1967).

In ORAUT 2014a, NIOSH appears to include the uranium bioassay results from these identified 

workers that were exposed in the powder room. In Neton 2014b, NIOSH observes: 

...at a number of sites different classes of workers during the same time period 

may have had monitoring programs that were conducted for different purposes. 

For example, construction and building trade workers, who worked intermittently 

in radiological areas, may have been monitored only when an incident was 

expected to have occurred, while those employees involved in routine process 

operations would have been routinely monitored on a frequency commensurate 

with their exposure potential. In this case, it would not be appropriate to combine 

the monitoring data for these two groups of workers into a single coworker model 
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that assumes a chronic exposure pattern. Rather, the default in this case should 

be to consider separate coworker models. (Neton 2014b)

In the case of the workers that were exposed to U-Alx, the frequency of monitoring increased 

when workers were found contaminated. In addition, workers were exposed to uranium 

compounds with an AMAD of 1 μm instead of the generally assumed 5 μm AMAD. In addition, 

the compounds of U-Alx handled in the powder room during this period showed a distinctly 

different excretion rate pattern than what one would expect for the classical solubility Types F, 

M, and S uranium compounds. As explained by Leggett et al. 2005, several months after the 

start of the U-Alx fuel fabrication program, it became evident from available monitoring data that 

the behavior of inhaled U-Alx differed from that of other forms of uranium that had been handled 

at this facility, and intake and lung retention were being underestimated based on the standard 

models and assumptions. For example, it was observed that the urinary excretion rate actually 

began to increase after workers ceased their internal exposure potential to radioactive materials. 

In these workers who had been removed from exposure, the rate of urinary excretion of uranium 

would continue to rise over a period of months before reaching a peak excretion rate, which was 

followed by a sharp decline. This pattern of excretion rate was observed by SC&A in the 

coworker data for workers who were listed as being restricted from work in the Powder Building 

(Atomics International 1967).

ORAUT 2014a uranium intake rates were aggregated during these periods: 1965–1968, 1969–

1972, 1973–1979, and 1980–1988. The 1965–1968 period of time is not appropriate for 

aggregating the results from workers exposed to uranium aluminide, as many high uranium 

excretion results were obtained in a period of time when the workers were restricted from work 

in areas subject to potential airborne radioactive material. Additionally, “less than detectable” or 

“zero” samples may have been included in the coworker model, which would not accurately 

reflect exposure to U-Alx, owing to the non-monotonic nature of the associated excretion pattern. 

Finding 13: NIOSH should discuss how uranium samples taken from workers in the U-Alx

powder operations are interpreted in the coworker analysis, and whether a separate assessment of 

coworker doses in the Powder Building may be warranted. In the latter case, it would be 

necessary to establish with sufficient accuracy which workers were potentially exposed to 

uranium aluminide in the Powder Building during the relevant period. 

5.4 OCCUPATIONS MONITORED FOR URANIUM INTAKE 

Occupational information was analyzed by cross-referencing dose records that specify an 

analytical method for detecting uranium with occupational information developed as described in 

Section 2.0. From Section 2.0, SC&A determined that the monitoring program appeared to 

target the workers most at risk (technical staff and mechanics) while also covering a crosssection of other occupations that could also be exposed. Table 26 shows the number and 

percentage of workers in each occupation monitored for uranium intakes compared to the total 

number of workers monitored for uranium. 

Based on Table 26, it appears that monitoring for uranium among the different occupations 

mirrors the overall site data presented in Section 1.0, with the Mechanics (38%) and Technical 

Staff (27%) receiving the most uranium monitoring while other, but not all, occupations were 
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monitored. These two occupation classes were monitored throughout SSFL’s operational history 

for uranium intakes, except there are no data available from the early-to-mid 1980s, as was noted

in Finding 11. Other occupations were intermittently monitored as would be expected. 

Table 26. Occupations Monitored for Uranium Intakes at SSFLa

Occupational Group Number of 

Individualsb Years Monitoring Data Available

Percentage of 

Individuals in an 

Occupational Group

Construction [redacted] 1965–1966, 1968–1970 1%

Electrician [redacted] 1965, 1971–1981, 1983 3%

Fireman [redacted] 1965–1970 2%

Health and Safety/

Health Physicsc

11 1965–1976, 1978–1980, 1983, 1988 5%

Inspector [redacted] 1975–1977, 1981, 1988 1%

Janitor 11 1965–1983 5%

Machinist [redacted] 1965–1968, 1970–1976, 1978, 

1980–1983

3%

Maintenanced14 1965–1976, 1978–1983 6%

Management [redacted] 1965–1983 3%

Mechanice62 1965–1982, 1986–1989 26%

Office Personnel 11 1965–1968, 1970, 1975–1980, 

1982–1983, 1985–1988

5%

Operator 0 1965–1976 0%

Quality Assurance [redacted] 1970–1981, 1983 <1%

Radiographer [redacted] 1975–1978, 1981–1982 <1%

Reactor Operator [redacted] 1965–1967, 1970, 1983 2%

Security [redacted] 1965–1970, 1986 3%

Shipping [redacted] 1965, 1966, 1968 <1%

Technical Stafff 63 1965–1982, 1986–1988 27%

Technician [redacted] 1967–1968, 1987 - 1988 1%

Truck Driver [redacted] 1973–1976, 1979, 1982 <1%

Unknowng[redacted] 1965–1966, 1971–1974, 1986–1988 2%

Welder [redacted] 1965–1969, 1980–1982 1%

Total 235 - 95% h

a

 Excludes dose records prior to 1965 and between January 1991 and mid-June 1993.

b

 An individual may be represented more than once if their occupation changed over the course of their 

monitoring. 

c

Includes medical staff, and health physicists and their support staff

d

 Includes individuals with ‘maintenance’ as part of their job title, utility services staff, and painters

e

 Includes individuals with ‘mechanic’ somewhere in their job title, and those who describe themselves as 

reactor and reactor fuel assemblers/disassemblers.

f

 Includes individuals with ‘engineer’ as part of their job title, physicists, chemists, and analysts

g

 Either a job title was not specified and marked as unknown, or it is unclear what the job title represents (for 

example, ‘Y&S’, ‘FMS’).

h

 Total percentage does not equal 100% due to rounding.

One exposure aspect not discussed in the internal coworker model is modeling for incidents. In 

the case of uranium, uranium fires were not uncommon occurrences. In a review of the SRDB, 

SC&A found 53 incidents involving uranium fires. It must be noted that most, but not all, of 

these documented fires occurred prior to 1965 and only two of the incidents appeared to result in 

bioassays being performed (post-1965). Nonetheless, incidents involving uranium fires in the 
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period of interest should be addressed. A summary of the incidents can be found in 

Attachment 2. 

Finding 14: The internal coworker model should specify the types of incidents that are assumed 

to be covered by the model, and when incidents would be expected to be reviewed separately 

with an incident-specific model. 
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6.0 EVALUATION OF NEED FOR MODELING OTHER 

CONSTITUENTS 

The internal dosimetry model addresses uranium, plutonium, and mixed fission products, as 

analyzed in the previous sections of this report, but is silent regarding other radionuclides 

potentially inhaled and/or ingested by workers. These radionuclides include isotopes of thorium, 

cesium, strontium, americium, polonium, and cobalt, as well as tritium. All but one of these 

radionuclides cannot be detected using the monitoring methods for uranium, plutonium, or mixed 

fission products. 

A number of research and production activities are associated with thorium and tritium. Cesium 

and strontium are mixed fission products that would be important contributors to dose. Some 

machining of cobalt may have occurred. While exposures to cobalt generated as a fission 

product from a reactor would normally be monitored using analytical methods associated with 

mixed fission products, it is possible that this monitoring approach was not used for the unique 

activity of machining unsealed cobalt, and it may be necessary to look at its isotopes separately 

when modeling exposures to workers found to have performed these activities. 

Bioassays for all of these radionuclides were specifically performed at some point during SSFL 

operations and cleanup, which may lend credence to their importance in determining radiation 

dose to workers. 

Isotopes associated with cesium, strontium, europium, and promethium can generally be 

estimated using activity ratios described in ORAUT-OTIB-0054 Rev. 02 (ORAUT 2014b) and 

do not need to be separately addressed in dose reconstructions. However, no mention is made of 

this in ORAUT-OTIB-0080. NIOSH should additionally demonstrate the feasibility of 

reconstructing exposures to thorium, tritium, and americium.

Finding 15: The internal dose coworker model is incomplete in that it does not address other 

radionuclides that were present on site and that could have substantively contributed to worker 

intakes. NIOSH should ascertain their ability to reconstruct intakes of thorium, tritium, 

americium, cesium, strontium, polonium, and cobalt.

More detailed information is provided below regarding how each of these isotopes was used or 

was present at SSFL, and to what extent, based on the available dose records. Information was 

taken from the site profile unless otherwise referenced. 

Thorium 

Facility Operations 

The Sodium Reactor Experiment (SRE) reactor, which operated from 1957 to 1964, used two 

cores over its lifetime, the second core containing thorium. The thorium fuel was manufactured 

in the Engineering Test Building (Building 4003). The operation consisted of loading of 

uranium and thorium metal slugs into metal tubes, filling the interstices with sodium metal, and 
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sealing the tubes (Rucker 2009, ORAUT 2006a). The fuel loaded in 1960 was 7.6% (weight) 

Th-232 with 93% enriched uranium (EU) (ORAUT 2010). 

The other reactor that used fuel containing thorium was the Advanced Epithermal Thorium 

Reactor (AETR), which operated from 1960 to 1974. The AETR was built to study and test 

reactor core configurations for thorium- and uranium-fuel reactors. The AETR facility was 

expanded in 1961 to include the Fast Critical Experiment Lab (Building 4100) for epithermal and 

fast neutron criticality tests, in which 20 different reactor core configurations were studied. The 

AETR’s first 9 core configurations (through 1965) contained various amounts of U-233 and 

Th-232 and were driven by 93% EU fuel. The AETR and Fast Critical Experiment Lab 

supported the development of reactors for the Southwest Atomic Power Association. For both 

the SRE and AETR, the thorium isotopes present in the fuel would have consisted mainly of 

Th-232 and its decay progeny Th-228 in partial equilibrium. Thorium-232 is also an activation 

product that was generated from operation of the reactors at SSFL (Rucker 2009, ORAUT 

2006a, ORAUT 2010). 

Research activities were conducted at SSFL regarding the reprocessing of used nuclear fuel. 

Rockwell International developed a process to effect a partial separation of used fuel, removing 

part of the mixed fission products, so that the material could be used again as reactor fuel. Tests 

were performed in a well-shielded Hot Cave in the Engineering Test Building. These 

experiments used up to one-kilogram quantities of un-irradiated uranium and thorium, and up to 

100-gram quantities of highly irradiated materials (Rucker 2009). 

In 1979, about 540 thorium plates containing about 200 kg of thorium were chamfered, removing 

an estimated total of 150 grams of thorium using a surface grinder (Tuttle 1979a; Tuttle 1979b). 

An enclosure around the head of the grinder and a chip catcher were used to capture the release 

of material ground off the plates. A vacuum cleaner with HEPA filter was attached to the 

enclosure (Tuttle 1979a). However, an adequate air flow could not be maintained and wet 

grinding was used to complete the chamfering (Begley 1979a). This operation was conducted in 

Building 001. The machinist was required to wear rubber gloves, shoe covers, and coveralls, and 

was required to undergo urinalysis testing to monitor internal exposure to thorium (Tuttle 

1979a). The machinist was W.C. Woods, who was assisted by G. Iskiyan. The supervisor was 

D.W. Hebdon. The work was conducted from February 28 to March 12, 1979 (Begley 1979b). 

Exposures to Thorium 

No facility-specific or individual-specific information was found in the SRDB regarding the 

amount of exposure received by workers to thorium. In the unscreened electronic dose records 

database, there are 49 records associated with specific nuclide analyses for thorium, all from 

1970, 1971, and the 1990s. When matched with occupational data, it is found that “mechanics” 

was the only occupation class monitored for thorium. The machinist that performed the grinding 

operation summarized above and the supervisor are in the electronic dose record database, but 

not for the year the grinding operation was performed. The assistant is not in the database. 
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Incidents Related to Thorium 

No incidents involving thorium were found in the SRDB or the DOE’s Occupational Reporting 

and Processing System, which contains incident reports since 1990. 

Tritium 

A safety review report was issued in 1992 that summarizes the uses and generation of tritium at 

SSFL (Tuttle 1992). Small amounts of tritium were formed by the reactors. Most of the tritium 

generated by the SSFL reactors remained bound within the reactor fuel because most of the fuel 

was metallic, or within the coolant; this is especially true of the SRE, where the coolant was a 

sodium-potassium compound. The tritium found in the groundwater at SSFL was most likely 

generated due to the interaction of neutrons in the reactor biological shielding (concrete) and 

with naturally occurring lithium within the molecular structure of the shielding. Tritium would 

then leach from the shielding into the surrounding groundwater. 

There were five other activities where tritium was used. 

1. Several thousand curies of tritium were used at the Hot Lab from 1964 to 1967 to 

investigate the properties of uranium-zirconium hydride systems. This tritium would 

have been used in small quantities and released through the Hot Lab exhaust system as a 

gas and not as tritiated water. 

2. Titanium and zirconium targets containing 1 to 120 Ci of tritium were used in a variety of 

neutron generators to produce high-energy neutrons for research and development 

projects. Small amounts of tritium outgas during use of these targets were to be trapped 

by oil in vacuum pumps or discharged to the atmosphere. 

3. Gas chromatographs were used for extremely sensitive analysis of organic compounds in 

water and other materials. An important part of the instrument is the detector, which, in 

some models, uses a titanium-tritium foil of roughly 0.2 Ci as a source of electrons. 

Essentially no tritium is released from these foils. 

4. A high-powered laser using a deuterium-fluorine reaction was operated at the Laser 

Experimental Test Facility (LETF) in Area I. The deuterium gas, furnished by Oak 

Ridge National Laboratory (ORNL), contained a small amount of tritium as an impurity, 

and this was released to the atmosphere during operation of the laser. 

5. Based on the SSFL Site Profile Site Description (ORAUT 2006a), reactor-generated 

tritium targets were bombarded by a Van de Graaff generator to produce neutrons. These 

accelerators were removed in 1962. 

There are 429 records in the unscreened electronic dose database where tritium was specifically 

analyzed. Health and safety personnel, mechanics, and technical staff were occupations that 

were monitored for tritium. 
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Americium 

Small quantities of americium (4 grams) were used as part of the Transuranic Management by 

Propartitioning-Separation (TRUMP-S) Program. This material was stored in the Fuel Storage 

Facility and was ultimately shipped offsite. Americium was generated as a byproduct of the use 

of plutonium at SSFL and would be a potential source of exposure in conjunction with 

plutonium. 

There are 24 records in the unscreened electronic dose database where americium was 

specifically analyzed from the late 1960s through 1993. Health and safety personnel, mechanics, 

an electrician and an inspector were occupations that were monitored for americium. 

Cesium 

Cesium was generated by the SSFL reactors and criticality test facilities as a fission product. 

The Gamma Irradiation Facility (GIF) used Cs-137 sources for radiation hardening tests of 

electrical components and food irradiation research. Cesium was present in nuclear support 

facilities like fuel disassembly and examination facilities, such as the Hot Lab and the 

Radioactive Materials Disposal Facility (RMDF), and waste management facilities. Cesium was 

also an environmental contaminant at the SSFL in the burn pit and the RMDF, and two small 

areas of contamination (Rockwell 1996) were found and remediated near the Fuel Storage 

Facility (referred to as T064 in the reference). 

There are 26 records in the unscreened electronic dose database where cesium was specifically 

analyzed, all in the 1990s. Mechanics was the primary occupation monitored for cesium, 

although one construction worker was also monitored. 

Strontium 

Strontium was generated by the SSFL reactors and criticality test facilities as a fission product. 

Strontium was present in nuclear support facilities, such as fuel disassembly and examination 

facilities and waste management facilities. Strontium-90 sealed sources were used at SSFL. 

Strontium contamination was found at the RMDF and it was a primary radiological contaminant 

at the burn pit. 

There are 68 records in the unscreened electronic dose database where strontium was specifically 

analyzed, from the 1960s through the 1990s. Mechanics was the primary occupation monitored 

for strontium, although one construction worker was also monitored, which is consistent with 

cesium as would be expected. 

Polonium 

A Polonium-Beryllium neutron source was used at SSFL (Heine 1970). The prevalence of 

polonium at SSFL, however, would have been primarily because it is in the uranium-238 decay 

chain. There are five records in the unscreened electronic dose database from 1966 for one 

worker where polonium was specifically analyzed. The worker’s occupation is unknown. 
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Cobalt 

Cobalt was not listed in SC&A’s comments in Issue 10 of the September 14 issues matrix 

update. Cobalt is a fission product from reactor operations that would have been monitored as 

such. Cobalt did have specific uses at SSFL. Cobalt-60 sources were used in the GIF for 

material and food irradiation research (ORAUT 2006a). In addition, “the Hot Lab was also used 

for manufacturing sealed sources, for leak checks on sources, and for cutting and machining 

operations that involved cobalt-60” (ORAUT 2009). It is not known whether the workers in the 

Hot Lab would have been monitored specifically for cobalt-60 during these cutting and 

machining operations or if they would have generally been monitored for mixed fission products. 

There are 18 records in the unscreened electronic dose database from 1986 and 1993 where 

cobalt was specifically analyzed. The primary occupation monitored was mechanics.
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REFERENCES 

Alexander 1967. Radiation Safety Unit Weekly Newsletter for Period Ending July 29, 1967. 

R.E. Alexander. Atomics International, Canoga Park, California, August 3, 1967. 

Arno 2013. SSFL Coworker Data Evaluation Rules – Prepared by Matthew Arno. Matthew 

Arno. Oak Ridge Associated Universities Team, Cincinnati, Ohio. June 13, 2013. 

Atomics International. Report of Inhalation Exposures at Atomics International, Letter to L.D. 

Low – U.S. Atomic Energy Commission, Document 6TAT-4446, August 1967.

Barnes, J.G., 2000. Results of Dose Screening/Assessment for Bioassay Samples 1989 – 1999, 

The Boeing Company, No. 090012, May 19, 2000 (Revised 12 February 2001). 

Baurmash 1967. Particle Size of UO2 Aerosol in Processing Room in Bldg. 1. Internal letter 

from L. Baurmash to A.N. Tschaeche. Atomics International, Canoga Park, California. 

November 27, 1967. 

Begley, F.E., 1979a. Internal letter from F.E. Begley to R.J. Tuttle, Change in the Thorium 

Grinding Operation Use Authorization 120. March 9, 1979. 

Begley, F.E., 1979b. Internal letter from F.E. Begley to R.J. Tuttle, Thorium Grinding 

Operation Use Authorization 120, March 20, 1979. 

Heine, W.F., 1970. Letter from W.F. Heine to D.D. Honey of the California State Department of 

Public Health, Re: Encapsulation of 1000 Ci Po-Be Neutron Source. November 13, 1970. 

SRDB 78236. 

ICRP 1994. Human Respiratory Tract Model for Radiological Protection. ICRP Publication 66.

Ann. ICRP 24 (1-3). 

Leggett, R.W., K.F. Eckerman, and J.D. Boice Jr., 2005. A Respiratory Model for Uranium 

Aluminide Based on Occupational Data. Institute of Physics Publishing – Journal of 

Radiological Protection. J. Radiol. Prot. 25 (2005) 405-416. December 6, 2005. 

Moore, J.D., 1980a. Letter from J.D. Moore to M.E. Remley, Quarterly Review of NMDF 

(T055) for Radiation Safety Second Calendar Quarter, 1980, Rockwell International. August 18, 

1980.

Moore, J.D., 1980b. Letter from J.D. Moore to M.E. Remley, Quarterly Review of NMDF 

(T055) for Radiation Safety Third Calendar Quarter, 1980, Rockwell International. December 

2, 1980. 

Moore, J.D., 1981a. Letter from J.D. Moore to M.E. Remley, Quarterly Review of NMDF 

(T055) for Radiation Safety Fourth Calendar Quarter, 1980, Rockwell International. March 5, 

1981.
Effective Date:

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However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

Moore, J.D., 1981b. Letter from J.D. Moore to M.E. Remley, Quarterly Review of NMDF 

(T055) for Radiation Safety Second Calendar Quarter, 1981, Rockwell International. August 13, 

1981.

Moore, J.D., 1982a. Letter from J.D. Moore to M.E. Remley, Quarterly Review of NMDF 

(T055) for Radiation Safety Second Calendar Quarter, 1982, Rockwell International. September 

8, 1982. 

Moore, J.D., 1982b. Letter from J.D. Moore to M.E. Remley, Quarterly Review of NMDF 

(T055) for Radiation Safety Third Calendar Quarter, 1982, Rockwell International. December 

2, 1982. 

Neton J.W., 2014a. Draft Criteria for the Evaluation and Use of Internal Exposure Coworker 

Datasets. J.W. Neton, Ph.D., CHP. Division of Compensation Analysis and Support. NIOSH. 

June 17, 2014. 

Neton, J.W., 2014b. Draft Criteria for the Evaluation and Use of Coworker Datasets – Rev. 02, 

J.W. Neton, Ph.D., CHP. Division of Compensation Analysis and Support, NIOSH. September 

30, 2014. 

ORAUT 2006a. Energy Technology Engineering Center – Site Description, ORAUT-TKBS0038-2, Rev. 00. Oak Ridge Associated Universities Team, Cincinnati, Ohio. February 2, 2006. 

ORAUT 2006b. Occupational Onsite Ambient Dose Reconstruction for DOE Sites, ORAUTPROC-0060, Rev. 01. Oak Ridge Associated Universities Team, Cincinnati, Ohio. June 28, 

2006.

ORAUT 2009. SEC Petition Evaluation Report for Petition SEC-00093, Santa Susana Field 

Laboratory-Area IV, SEC-00093, Rev. 00, Oak Ridge Associated Universities Team, Cincinnati, 

Ohio. April 28, 2009. 

ORAUT 2010. Area IV of the Santa Susana Field Laboratory, the Canoga Avenue Facility, the 

Downey Facility, and the De Soto Avenue Facility (sometimes referred to as the Energy 

Technology Engineering Center [ETEC] or Atomics International) – Occupational Internal 

Dose. ORAUT-TKBS-0038-5, Rev. 01. Oak Ridge Associated Universities Team, Cincinnati, 

Ohio. April 26, 2010. 

ORAUT 2012. Analysis of Stratified Coworker Datasets. ORAUT-RPRT-0053, Rev. 01. Oak 

Ridge Associated Universities Team, Cincinnati, Ohio. July 16, 2012. 

ORAUT 2014a. Internal Coworker Dosimetry Data for Area IV of the Santa Susana Laboratory 

and the De Soto Avenue Facility. ORAUT-OTIB-0080, Rev. 00. Oak Ridge Associated 

Universities Team, Cincinnati, Ohio. March 14, 2014. 
Effective Date:

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However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

ORAUT 2014b. Fission and Activation Product Assignment for Internal Dose-Related Gross 

Beta and Gross Gamma Analyses. ORAUT-OTIB-0054, Rev. 02. Oak Ridge Associated 

Universities Team, Cincinnati, Ohio. March 6, 2014. 

Rockwell 1975. Special Nuclear Material Control Program for Plutonium-Uranium Used by 

Atomics International Division of Rockwell International in Research and Development 

Activities Under Special Nuclear Material License No. SNM-21, Rockwell International. July 

29, 1975. 

Rockwell 1986. Annual Review of Radiological Controls – 1983, Rockwell International. 

January 25, 1986. 

Rockwell 1996. Rocketdyne Division Annual Site Environmental Report Santa Susana Field 

Laboratory and DeSoto Sites 1995, RI/RD96-140, Rockwell Aerospace/Rocketdyne. SRDB 

19123, July 30, 1996. 

Singh et al. 2010. ProUCL Version 4.1.00 Technical Guide (Draft). EPA/600/R-07/041, U.S. 

Environmental Protection Agency, Office of Research and Development, Washington, DC. May 

2010.

Rucker, T.L., 2009. Radionuclides Related to Historical Operations at the Santa Susana Field 

Laboratory Area IV, Science Applications International Corporation. March 2009. 

Saxe 1967. Report of Inhalation Exposures at Atomics International. D. Saxe letter to Lawrence 

D. Low. Atomics International, Canoga Park, California. August 15, 1967. 

Tuttle, R.J., 1979a. Internal letter from R.J. Tuttle to Isotopes Committee, Review of Proposal to 

Machine Thorium. February 14, 1979. 

Tuttle, R.J., 1979b. Internal letter from R.J. Tuttle to Isotopes Committee, Authorization for Use 

of Radioactive Materials or Radiation Producing Devices, Authorization No. 120. February 19, 

1979. 

Tuttle, R.J., 1992. Tritium Production and Release to Groundwater at SSFL, Safety Review 

Report, RI/RD92-186, Rockwell International. December 1, 1992. SRDB 19158.
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Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

ATTACHMENT 1: BREAKDOWN OF INTERNAL MONITORING 

ENTRIES CONTAINED IN THE “NUCLIDE/ANALYSIS” COLUMN OF 

THE COWORKER DATABASE 

Nuclide/Analysis

Entry

Total Entries

(% of Total)

Nuclide/Analysis

Entry

Total Entries

(% of Total)

Nuclide/Analysis

Entry

Total Entries

(% of Total)

UR 12,883 (34.519%) Tot U 11 (0.029%) alpha, beta 3 (0.008%)

UF 7,593 (20.345%) U-Tot 11 (0.029%) GB Beta 3 (0.008%)

FP 6,004 (16.087%) U-238 9 (0.024%) Iso U 3 (0.008%)

MFP 4,735 (12.687%) Ca-45 8 (0.021%) MFP(G) 3 (0.008%)

Blank or “XX” 1,322 (3.542%) UF 8 (0.021%) alpha, beta, gamma 2 (0.005%)

GA 880 (2.358%) Ba-140 7 (0.019%) B 2 (0.005%)

GB 745 (1.996%) beta gamma 7 (0.019%) GB 2 (0.005%)

PUA 644 (1.726%) Ce-141 7 (0.019%) HG 2 (0.005%)

PU 534 (1.431%) Ce-144 7 (0.019%) M 2 (0.005%)

H-3 429 (1.149%) Cs-134 7 (0.019%) UFP 2 (0.005%)

GB(H) 167 (0.447%) Fe-59 7 (0.019%) GB Beta 1 (0.003%)

Alpha 148 (0.397%) Gamma 7 (0.019%) EU 1 (0.003%)

Beta-gamma 121 (0.324%) GB 7 (0.019%) FP-3AG-Co-60 1 (0.003%)

GB(L) 106 (0.284%) Gross beta (L) 7 (0.019%) G 1 (0.003%)

Beta 103 (0.276%) Mn-54 7 (0.019%) GAL 1 (0.003%)

AP 100 (0.268%) Ra-226 7 (0.019%) gamma, beta 1 (0.003%)

Gamma 88 (0.236%) Ru-103 7 (0.019%) GAZ 1 (0.003%)

Sr-90 49 (0.131%) Ru-106 7 (0.019%) GAZA 1 (0.003%)

U 32 (0.086%) Zn-65 7 (0.019%) GI 1 (0.003%)

U-ISO 32 (0.086%) Co-58 6 (0.016%) GR-B-K-39 1 (0.003%)

Th 28 (0.075%) GR-B-K-40 6 (0.016%) GR-B-K40 1 (0.003%)

Am-241 24 (0.064%) Zr-95 6 (0.016%) Gross -Beta 1 (0.003%)

MFPG 23 (0.062%) Cm-242 5 (0.013%) K 1 (0.003%)

U-total 22 (0.059%) Cm-244 5 (0.013%) MFP 1 (0.003%)

SR-90 (D) 21 (0.056%) Gross alpha 5 (0.013%) MFP,UR 1 (0.003%)

Cs-137 20 (0.054%) LowBeta 5 (0.013%) NDA 1 (0.003%)

MFPB 20 (0.054%) MFP(B) 5 (0.013%) P 1 (0.003%)

Pm-147 18 (0.048%) Midbeta 5 (0.013%) PA 1 (0.003%)

CFP 16 (0.043%) Np-237 5 (0.013%) Pu UF 1 (0.003%)

GRB(-K40) 14 (0.038%) Po 5 (0.013%) Pu/MFP 1 (0.003%)

U-234 13 (0.035%) Pu-242 5 (0.013%) Pu238/239 1 (0.003%)

Co-60 12 (0.032%) Th-230 5 (0.013%) Pu-238/239 1 (0.003%)

I-131 12 (0.032%) Th-232 5 (0.013%) Pu-238-239 1 (0.003%)

Th-228 12 (0.032%) Topbeta 5 (0.013%) Sr 1 (0.003%)

U-235 12 (0.032%) Be-7 4 (0.011%) Sr 90 1 (0.003%)

Ac 11 (0.029%) GB-K40 4 (0.011%) Tri 1 (0.003%)

K-40 11 (0.029%) GB-K-40 4 (0.011%) U, GB 1 (0.003%)

Pu-238 11 (0.029%) Pu-23 4 (0.011%) UR 1 (0.003%)

Pu-239 11 (0.029%) UISO 4 (0.011%) UTOT 1 (0.003%)
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ATTACHMENT 2: LIST OF URANIUM FIRE INCIDENTS 

The following table describes instances of uranium fires that have occurred at SSFL, as found in the SRDB. 

SRDB 

Number Date

Number of 

people 

involved

Location

Number of 

Bioassays 

post-1964

Description

39492 6/30/1960 Room 426-13 -

1 gallon can containing uranium turnings immersed in watersoluble machining coolant.

39493 1/10/1978 [redacted] Building 028 -

Small uranium fire. Small amount of contamination on lip of arc 

furnace measuring 60 dpm/100 cm2alpha. Full face respiratory 

was being worn.

39494 10/26/1982 [redacted] ATR area, Room 119-36 -

UALx fire, 1888 grams U-235. Powder ignited as it was being 

poured into a blender in a glovebox. No personal contamination 

survey performed.

39496 11/2/1960 [redacted] 001, Room 1155 -

Magnesium flashes while opening induction furnace. 100 kg of 

1.945 enriched uranium, daughter products, and oxides. 

Respirators worn. Bioassays neg. 8.37E-10 μCi/cc beta gamma 

air sample.

39497 11/30/1960 [redacted] Pluto area #1244 -

279 grams 93% UX in hydrogen sintering furnace. No abnormal 

air contamination. Operator smeared. No bioassay.

39498 12/3/1966 HQ1, 111D-62 - File contains air sampling results

39499 6/20/1960 [redacted] HQ1, 1063 -

2,300 grams of 10% enriched uranium not handled correctly, 

creating U3O8 which was put in a can. Hydrogen buildup blew up 

can. No bioassays. 

39500 12/19/1963 [redacted] Building 4, 424-68 -

40-50 grams of uranium carbide involved with 30 grams being

recovered. Smears on 3 personnel showed no contamination.

39501 11/2/1961

793-1416,

793-226,

793-413

Fed area, Room 1026 -

1 gram of UO2 powder. Personnel contaminated with 2 receiving 

bioassays. 500 dpm alpha and 1200 dpm beta-gamma 

contamination.

39502 3/20/1962

788-10

752-

[redacted]

779-28

FEPF, 1253 -

3 16% enriched U-Mo Hallam slugs. 156 grams UX, 4,300 grams 

U. Surface contamination 500 dpm alpha, 1533 dpm beta+F12-

gamma.

39503 8/21/1951 Downey Room A-126 - Uranium chips caught fire

39504 12/10/1966 HQ, Bldg 004 -

Two fires. One was a mixture of Na and Cs-137. Other was 

uranium metal chips. No significant inhalation exposures.

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SRDB 

Number Date

Number of 

people 

involved

Location

Number of 

Bioassays 

post-1964

Description

39505 2/13/1967 UCPFF -

Small amount of UC in a metal exhaust duct ignited. No 

unfiltered release occurred. 

39506 1/30/1967 UCPFF -

Uranium fire in retention tanks of vacuum system. Ducts burned 

through but no large scale release and no personnel exposure. 

Nasal smears taken of 15 men. 1 fireman had contaminated 

clothing.

39507 1/27/1967 Aisle 1110 and Room 110-65 - JAERI fuel (20% uranium). No internal exposure.

39508 3/10/1967 -

Uranium metal chips ignited from sparks from cutting metal duct. 

Fireman and maintenance man scheduled for bioassays due to 

contamination on face and hair. 

39509 5/21/1967 On truck -

Explosion in a 55-gal drum containing 1-gal cans of UC sludge. 

Vermiculite and uranium contamination 900 to 5,000 dpm/

100 cm2beta.

39510 5/21/1967 RMDF storage yard -

55-gal drum containing uranium metal under CaCO3 powder

discovered burning. Air samples, nasal swabs and contamination

surveys revealed no significant release or contamination.

39511 6/13/1970 [redacted] HQ001, Room 119-42 5

Oven drying of 20 lbs of uranium chips that caught fire. Floor 

contaminated <50 dpm/100cm2alpha and 30 dpm/100cm2beta. 

Airborne was 1.6E-10 μCi/cc. Bioassays taken. All beta-gamma 

below dpm/100cm2of nasal, clothing, skin. Highest alpha was 

15 dpm/100cm2nasal, 20 dpm/100cm2clothing, 28 dpm/100cm2

skin.

39512 8/26/1976 [redacted] SNM Vault weigh room 5

Sealed transfer can of uranium fines was dropped, causing 

autoignition. Floor contaminated with 200 dpm/100cm2alpha. 

One person suffered finger cut with 40 dpm alpha which was 

deconned. Nasal smears indicated max of 11 dpm alpha. 

39513 4/27/1977 [redacted] ATR-QA lab -

Fire in lab. Substances burned not specified. All nasal, hands, 

shoes, socks below 20 dpm alpha. No internal or external 

exposures.

39514 1/9/1978

[redacted]

and several 

others

ATR Powder Room -

UALx fire in crusher glove box. All smears on glove box 

<20 dpm alpha.

39515 9/14/1965 [redacted] Bldg 1, SS Vault Pad -

Fire in 1-gal can of DU from ESADA Project. No contamination 

found and nasal smears negative.

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SRDB 

Number Date

Number of 

people 

involved

Location

Number of 

Bioassays 

post-1964

Description

39516 6/12/1965 [redacted] FED II filter plenum -

Involved uranium carbide. Significant contamination near filter 

bank. Minor personnel contamination noted. One individual had 

up to 25 dpm alpha nasal. Up to 155 dpm alpha hair, up to 

675 dpm alpha trousers, up to 300 dpm alpha skin, up to 390 dpm 

alpha shirt, up to 500 dpm alpha shoes.

39517 6/26/1964 [redacted] Canoga 1, Room 1126 -

4.91% enriched UC cuttings dispersed during vacuuming. Five 

people contaminated. Up to 64 dpm alpha and 22 dpm betagamma nasal, 50 dpm alpha and 25 dpm beta-gamma body (face, 

hair, hands, arms).

39518 6/24/1964 [redacted] Building 1, Room 2021 -

1 gram UC on fire in desiccator. One individual had 4 dpm alpha 

and 15 dpm beta-gamma nasal. Shoes had 30 dpm alpha.

39519 2/13/1964 [redacted] Bldg 1, Room 1126 -

Fire in vacuum cleaner with unknown amount of 4.9% enriched 

UC. One individual received first aid. Breathing apparatus worn.

Air samples indicated 1.8E-11 μCi/cc alpha and 8.4e-10 μCi/cc 

beta-gamma. Floor smears showed 1,000 dpm alpha and 

1,400 dpm beta-gamma.

39520 9/10/1959 [redacted] Room 807 -

Two one-gal cans of U-Nor/Zirconium chips. 12 dpm betagamma on neck.

39521 1/6/1964 [redacted] Bldg 1, Room 1126 -

Fire in vacuum cleaner with unknown amount of UC. All 

received bioassays. Breathing apparatus worn. Air samples 

indicated 1.8E-11 μCi/cc alpha and 1.0e-10 μCi/cc beta-gamma. 

Floor smears showed 400 dpm alpha and 1000 dpm beta-gamma.

39522 3/4/1964 HQ-1, Rooms 1131, 1114 -

UC fire in ducting between rooms. Breathing apparatus worn. 

Air samples indicated 6.8E-11 μCi/cc alpha and 2.8e-10 μCi/cc

beta-gamma. Floor smears showed 1,000 dpm alpha and 

2,300 dpm beta-gamma.

39523 8/31/1959 [redacted] Room 901 - UC slug ignited in desicator.

59524 3/15/1961 [redacted] HQ-1, Room 1059 -

Up to 900 grams UC, 95.1% uranium, 3.02% enriched. Burned 

rubber pants.

59525 3/16/1961 [redacted] Room 1059, Cresta grinder - UC grindings. No increase in air activity.

59526 6/23/1961 [redacted] Room 1059 - 13 grams UC pellet caught fire.

59527 7/8/1967 UCPFF -

Peeled paint on retention tank indicated a fire occurred at one 

time.

--

--
Effective Date:

November 24, 2014

Revision No.

0 – Draft

Document Number:

SCA-TR-OTIB2014-0080

Page No.

Page 67 of 69

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

SRDB 

Number Date

Number of 

people 

involved

Location

Number of 

Bioassays 

post-1964

Description

39544 5/5/1961 [redacted] SS Vault, Room 1142 -

3.25% enriched UC. Explosion in can while being held. Dropped 

it and fire came out of can. Up to 18 dpm alpha and 18 dpm betagamma on nose, 2 dpm alpha and 30 dpm beta-gamma face, 

2 dpm alpha and 55 dpm beta-gamma on neck, and 12 dpm alpha 

in hair.

39545 7/29/1964 [redacted] Room 1131, Furnace 3 - Implosion of furnace. Up to 32 dpm alpha face and arms.

39546 2/22/1961 [redacted] Room 1059 - UC fire in grinding operation. 

39547 6/28/1976

[redacted], 

including 

[redacted]

ATR Powder Room -

3 kg of U-ALx briquet dropped. All received nasal swipes, with 

one receiving 20 dpm/100cm2. Up to 4,000 dpm/100cm2alpha 

found on clothing.

39548 7/5/1961 [redacted] Building 009, SanSu -

Two 1-gal containers containing 30 UC slugs. No personnel 

contamination.

39549 8/26/1964 [redacted] Room 1059 -

Small amount of normal and enriched UC caught fire in Cresta 

grinder. 

39574 9/23/1961 [redacted] SNAP, room 1273 - Fire while machining uranium on a lathe. 

39575 8/25/1964 [redacted] Room 1273 -

UC zirconium and UX material ignited. Smears and nasal smears 

all showed background.

39598 5/13/1965 [redacted] Bldg 022 AMF evaporator -

Overflow of chemically treated contaminated liquid. Received 

splashes to hard hat, lab coat, and goggles. No contamination 

found.

39599 8/21/1963 [redacted] SNAP, Room 1252 -

U-Zr fire - sandblasting unit. Breathing apparatus worn. Surface

contamination up to 3,500 dpm alpha and 4,200 dpm betagamma. Room 1256 also contaminated. Fireman's pants

confiscated. Nasal smears positive.

39609 8/9/1962 [redacted] Room 1261 -

Degreaser overheated causing fire and TCE fumes. 2 firemen 

responded, using self-contained breathing units. 

39617 8/20/1959 Room 807 - U-Zr shavings ignited.

39621 9/23/1964 [redacted] HQ-1, Room 1114 -

Lid on a one-quart paint can containing 5,366 grams of 4.91% 

enriched UC vacuum dust blew off. No personnel contamination. 

Surface smears showed 2,000 dpm/100cm2alpha and 2,400 dpm 

beta-gamma.

39622 2/27/1963 [redacted] Room 1059 -

Fire in R/A exhaust system. Air concentration of 3.4E-11 alpha 

and 5.0E-11 beta-gamma.

--
Effective Date:

November 24, 2014

Revision No.

0 – Draft

Document Number:

SCA-TR-OTIB2014-0080

Page No.

Page 68 of 69

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

SRDB 

Number Date

Number of 

people 

involved

Location

Number of 

Bioassays 

post-1964

Description

39623 10/1/1964 [redacted] HQ-1, Room 1126 -

Small amount of UC burned. Worker was not wearing shoe 

covers or lab coat. Shoes had 30 dpm alpha, pants 45 dpm alpha 

and 90 dpm beta-gamma.

39624 1/12/1961 [redacted] Bldg. 4, Room 423-29 -

3% enriched UC ignited while being welded. Nasal smear 

showed no significant contamination. Lab coat caught fire and 

was used to smother fire. Surface smears showed 300 dpm alpha 

and 500 dpm beta-gamma.

39625 10/3/1961 [redacted] Room 1063 -

1 pint of Hexane solution in a bread pan ignited from a spark 

created by striking 2 UC slugs against one another. 
Effective Date:

November 24, 2014

Revision No.

0 – Draft

Document Number:

SCA-TR-OTIB2014-0080

Page No.

Page 69 of 69

NOTICE: This report has been reviewed for Privacy Act information and has been cleared for distribution. 

However, this report is pre-decisional and has not been reviewed by the Advisory Board on Radiation and Worker 

Health for factual accuracy or applicability within the requirements of 42 CFR 82. 

ATTACHMENT 3: OCCUPATIONAL CATEGORIZATION 

The following table shows how occupational names were condensed down to a more generalized 

list. 

Construction

Construction Carpenter -- --

Mechanic

Mechanic Air Refrigeration 

Mechanic

Instrument Mechanic Instrumentation Mechanic

Lab mechanic/reactor 

assembly

Maintenance Mechanic Mechanic AER Mechanic AER Lab

Mechanic/Nuclear 

Materials Worker

Mechanic/Reactor 

Operator

Reactor Assembly Reactor Assemblyman

Reactor Fuel Disassembler Structure Mechanic -- --

Technical Staff

Analyst-Production Chemist Physicist Associate, Test 

Development

Engineer Engineering Lab Assistant Engineering Technician Lab Assistant

Metallurgical Engineer Research Analyst Sr Technical Specialist Tech for Nuclear Research

Test Engineer Eng. Office Personnel -- --

Health and Safety/Health Physics

Associate Health Physicist Health and Safety Rep Health Analyst Health Physicist

Health Rep Nurse -- --

Office Personnel

Office Personnel Clerk -- --

Operator

Chemical Operator Control Center Operator Sr. Production Control --

Electrician

Electrician Maintenance Electrician -- --

Management

Manager Management Analyst Manager Research Manufacturing Liaison

Quality Assurance

Manager – QA QA -- --

Machinist

Milling Machine Machinist Lathe Machinist Pipe Shop --

Maintenance

MTS III, IV, V Ops/Maintenance 

Technician

Plant Maintenance Painter

Utility Man – Machine 

Shop

Utility Man – General Plumber

--

Shipping

Ore Shipping -- -- --

Inspector

Inspector Reactor Assembly 

Inspector

ISI Inspector

--

Janitor

Janitor Custodian -- --

Radiography

X-ray Tech -- -- --