October 20, 2006 

Mr. David Staudt 

Centers for Disease Control and Prevention 

Acquisition and Assistance Field Branch 

Post Office Box 18070 

626 Cochrans Mill Road – B-140 

Pittsburgh, PA 15236-0295 

Subject: Contract No. 200-2004-03805, Task Order 1: Paducah Gaseous Diffusion Plant 

Site Profile Review, SCA-TR-TASK1-0016 

Dear Mr. Staudt: 

SC&A is pleased to submit to NIOSH and the Advisory Board our draft Paducah Gaseous 

Diffusion Plant Site Profile Review. Please note that Attachment 3: Summary of Expert 

Interviews, is currently being reviewed to confirm that no classified information was 

inadvertently included, and will be forwarded under separate cover when the review is 

completed. 

If you have any questions or comments on this report, please contact John Mauro at 732-530-

0104. We look forward to discussing this draft report with NIOSH and the Advisory Board. 

Project Manager 

cc: P. Ziemer, PhD, Board Chairperson A. Makhijani, PhD, SC&A 

Advisory Board Members H. Behling, SC&A 

L. Wade, PhD, NIOSH M. Thorne, SC&A 

L. Elliott, NIOSH H. Chmelynski, SC&A 

J. Neton, PhD, NIOSH J. Fitzgerald, Saliant 

S. Hinnefeld, NIOSH J. Lipsztein, SC&A 

L. Homoki-Titus, NIOSH K. Robertson-DeMers, CHP, Saliant 

A. Brand, NIOSH S. Ostrow, PhD, SC&A 

J. Broehm, NIOSH K. Behling, SC&A 

L. Shields, NIOSH T. Bell, Saliant 

Project File (ANIOS/001/16) 

Sincerely, 

John Mauro, PhD, CHP 

1608 SPRING HILL ROAD, SUITE 400 • VIENNA, VIRGINIA • 22182 • 703.893.6600 • FAX 703.821.8236 
 

Draft 

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH 

National Institute of Occupational Safety and Health 

Paducah Gaseous Diffusion Plant Site Profile Review 

Contract No. 200-2004-03805 

Task Order No. 1 

SCA-TR-TASK1-0016
 

Prepared by 

S. Cohen & Associates 

1608 Spring Hill Road, Suite 400 

Vienna, Virginia 22182 

Saliant, Inc.
 

5579 Catholic Church Road 

Jefferson, Maryland 21755
 

October 2006 

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 pre-decisional 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. 
 

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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-TASK1-0016 

Effective Date: 

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Revision No. 0 (Draft) 

PADUCAH GASEOUS DIFFUSION PLANT 

SITE PROFILE REVIEW Page 2 of 153 

Task Manager: 

________________________ Date: ___________ 

Christopher Nelson 

Supersedes: 

N/A 

Project Manager: 

________________________ Date: ___________ 

John Mauro, PhD, CHP 
 

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TABLE OF CONTENTS 

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

1.0	 Executive Summary .............................................................................................................9 

1.1	 Summary of Strengths............................................................................................11 

1.2	 Summary of Findings.............................................................................................11 

2.0	 Scope and Introduction ......................................................................................................17 

2.1	 Review Scope.........................................................................................................18 

2.2	 Review Approach...................................................................................................20 

2.3	 Report Organization...............................................................................................21 

3.0	 Assessment Criteria and Methods......................................................................................22 

3.1	 Objectives ..............................................................................................................22 

3.1.1	 Objective 1: Completeness of Data Sources .............................................22 

3.1.2	 Objective 2: Technical Accuracy..............................................................22 

3.1.3	 Objective 3: Adequacy of Data.................................................................23 

3.1.4	 Objective 4: Consistency Among Site Profiles.........................................23 

3.1.5	 Objective 5: Regulatory Compliance ........................................................23 

4.0	 Site Profile Strengths .........................................................................................................27 

4.1	 Completeness of Data ............................................................................................27 

4.2	 Adequacy of Data ..................................................................................................27 

4.3	 Technical Accuracy/Claimant Favorability ...........................................................28 

5.0	 Vertical Issues ....................................................................................................................29 

5.1	 Site Description TBD.............................................................................................29 

5.1.1	 Issue 1: The Enrichment Facilities at PGDP are no longer under 

DOE Regulation.........................................................................................29 

5.1.2	 Issue 2: The Enrichment Level at PGDP is no longer Limited to 

About 2.5% ................................................................................................29 

5.1.3	 Issue 3: The Site Map in the TBD is Inadequate ......................................30 

5.1.4	 Issue 4: The Material Processed to make Feed Material was not Just 

UO3 ............................................................................................................30 

5.1.5	 Issue 5: Average External Doses are biased Low by the Use of 

Assigned Zero Values................................................................................31 

5.1.6	 Issue 6: Incidents and Special Activities need to be Better 

Referenced .................................................................................................31 

5.1.7	 Issue 7: The TBD Needs to Consider Operations Other than 

Gaseous Diffusion......................................................................................31 

5.1.8	 Issue 8: The Intended Use of the Bechtel-Jacobs Radionuclide 

Concentrations is not Clear........................................................................31 

5.2	 Occupational Medical Dose ...................................................................................34 

5.2.1	 Issue 1: The TBD fails to Adequately Define and Assess 

Occupational Medical Exposure ................................................................34 
 

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5.2.2	 Issue 2: Techniques and Protocols Increase Uncertainty of DCFs 

listed in the TBD........................................................................................35 

5.2.3	 Issue 3: Frequency and Type of X-ray Exposure Is Uncertain .................36 

5.2.4	 Issue 4: Other Potential Medical Exposures have not been Identified .....37 

5.2.5	 Issue 5: There are Additional Factors that Contribute to 

Uncertainties ..............................................................................................37 

5.3	 Occupational Environmental Dose ........................................................................38 

5.3.1	 Issue 1: Onsite Environmental Exposures are based on Site 

Boundary Data ...........................................................................................38 

5.3.2	 Issue 2: There are No Comparisons between Predicted and 

Measured Environmental Exposures .........................................................39 

5.3.3	 Issue 3: The Pre-operational Measurements could not have detected 

Ambient Levels of Uranium in Air ............................................................39 

5.3.4	 Issue 4: The Ambient Air Sampling Collection Network was not 

Intended for Monitoring Onsite Concentrations........................................39 

5.3.5	 Issue 5: Environmental Dose Gaps Exist in the TBD ...............................39 

5.3.6	 Issue 6: Comments on Section 4.3.2, Radiological Conditions in 

Cylinder Storage Yards..............................................................................40 

5.3.7	 Issue 7: There is Little Characterization of the Effect of Plant 

Releases on the Onsite Levels of Exposure ...............................................40 

5.4	 Occupational Internal Dose....................................................................................40 

5.4.1	 Key Issues ..................................................................................................41 

5.4.2	 Section 5.2: “Source Term” ......................................................................46 

5.4.3	 Section 5.3: “In-Vitro Measurement Methods” ........................................59 

5.4.4	 Section 5.4: “In-Vivo Measurement Methods” ........................................64 

5.5	 Occupational External Dose ..................................................................................66 

5.5.1	 Issue 1: Questionable Assumptions for the Assessment of Recorded 

Skin Dose...................................................................................................69 

5.5.2	 Issue 2: Questionable Assumptions for Assigning Skin and Deep 

Dose for Unmonitored Workers Prior to 1960 by Means of 

Coworker Data...........................................................................................72 

5.5.3	 Issue 3: Issues related to ORAUT-OTIB-0031.........................................80 

5.5.4	 Issue 4: Issues Related to the Assessment of Neutron Exposures at 

PGDP .........................................................................................................81 

6.0	 Overall Adequacy of the Site Profile as a Basis for Dose Reconstruction ........................83 

6.1	 Satisfying the Five Objectives ...............................................................................83 

6.2	 Usability of Site Profile for Intended Purposes .....................................................90 

6.3 Unresolved Policy or Generic Technical Issues ....................................................91 

References......................................................................................................................................93 

Attachment 1: Richard Miller Letter to HHS and NIOSH .........................................................100 

Attachment 2: Technical Documents Considered during the Review ........................................112 

Attachment 3: Summary of Expert Interviews ...........................................................................113 

Attachment 4: Consistency among Site Profiles.........................................................................114 
 

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ACRONYMS AND ABBREVIATIONS 

Advisory Board Advisory Board on Radiation and Worker Health 

AEC Atomic Energy Commission 

Al Aluminium 

AMAD Aerodynamic Median Activity Diameter 

ATSDR Agency for Toxic Substance and Disease Registry 

AWE Atomic Weapons Employer 

BJC Bechtel Jacobs Corporation 

Bq Becquerel 

CDC Centers for Disease Control and Prevention 

CEDE Committed Effective Dose Equivalent 

CFR Code of Federal Regulations 

Ci Curie 

CIP Cascade Improvement Project 

COI Conflict of Interest 

CUP Cascade Upgrade Program 

DAC Derived Air Concentration 

DCF Dose Conversion Factor 

D&D Decontamination and Decommissioning 

DOL Department of Labor 

DR Dose Reconstruction 

DOE Department of Energy 

dpm Disintegrations per Minute 

DU Depleted Uranium 

FDA U.S. Food and Drug Administration 

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000 

EPA Environmental Protection Agency 

ERDA Energy Research and Development Agency 

ERT Enriched Reactor Tails 

ESE Entrance Skin Exposure 

EU Enriched Uranium 

FEMP Fernald Environmental Management Project 
 

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GM Geometric Means 

GSD Geometric Standard Deviation 

HEPA High Efficiency Particulate Air 

HHS Health and Human Services 

HVL Half Value Layer 

IAAP Iowa Army Ammunition Plant 

ICP-MS Inductively Coupled Plasma-Mass Spectrometry 

ICRP International Commission on Radiological Protection 

IMBA Integrated Modules for Bioassay Analysis 

IREP Interactive RadioEpidemiological Program 

keV Kilo electron Volt 

KPA Kinetic Phosphorescence Analysis 

kVp Kilovolt Potential 

LANL Los Alamos National Laboratory 

LAT Lateral 

LLNL Lawrence Livermore National Laboratory 

LOD Limit of Detection 

μCi Microcurie 

mA Milliampere 

MAC Maximum Allowable Concentration 

MCW Mallinckrodt Chemical Works 

MDA Minimum Detectable Activity 

MDC Minimum Detectable Concentration 

MDL Minimum Detectable Level 

MeV Mega-electron Volt 

NCRP National Council on Radiation Protection and Measurements 

NIOSH National Institute for Occupational Safety and Health 

NOD Net Optical Density 

NRC Nuclear Regulatory Commission 

NTA Eastman Kodak Nuclear Track Film Type A 

NTS Nevada Test Site 

NU Natural Uranium 
 

 

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OCAS Office of Compensation Analysis and Support 

ORAU Oak Ridge Associated Universities 

ORNL Oak Ridge National Laboratory 

OTIB ORAU Technical Information Bulletin 

PA Posterior-Anterior 

PACE Paper, Allied Industrial, Chemical and Energy Workers 

pCi pico curies 

PFG Photofluorography 

PGDP Paducah Gaseous Diffusion Plant 

PNNL Pacific Northwest National Laboratory 

POC Probability of Causation 

REM Roentgen equivalent man 

RFP Rocky Flats Plant 

RU Recycled Uranium 

SAIC Scientific Applications International Corporation 

SC&A S. Cohen and Associates 

SEC Special Exposure Cohort 

SID Source-to-image distance 

SNL Sandia National Laboratory 

SOP Standard Operating Procedure 

SRS Savannah River Site 

SSD Source-to-Skin Distance 

TBD Technical Basis Document 

TIB NIOSH Technical Information Bulletin 

TLD Thermoluminescent Dosimeter 

TLND Thermoluminescent neutron dosimeter 

TRU Transuranics 

UF4 Uranium Tetrafluoride 

UF6 Uranium Hexafluoride 

UO3 Uranium Trioxide 

U3O8 Urano-uranic Oxide 

UO2F2 Uranyl Fluoride 
 

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USEC United States Enrichment Corporation 

WFO Work For Others 

Y-12 Y-12 National Security Complex 
 

 

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1.0 EXECUTIVE SUMMARY 

This report provides the results of an independent audit conducted by S. Cohen and Associates 

(SC&A) of the technical basis documents (TBDs) that make up the site profile for the Paducah 

Gaseous Diffusion Plant (PGDP) developed by the National Institute of Occupational Safety and 

Health (NIOSH). This audit was conducted during the period April 2006–September 2006 in 

support of the Advisory Board on Radiation and Worker Health (Advisory Board) in the latter’s 

statutory responsibility under the Energy Employees Occupational Illness Compensation 

Program Act (EEOICPA) to conduct such reviews and to advise the Secretary of Health and 

Human Services (HHS) on the “completeness and adequacy” of the EEOICPA program. 

A subgroup of PGDP workers has been designated as a Special Exposure Cohort (SEC) as 

defined in 42 CFR Part 83. This designation allows those workers with a covered cancer to be 

considered for a compensation decision under the provisions of the EEOICPA without requiring 

a probability of causation (POC) calculation. Other workers with uncovered cancers would have 

to have such a calculation made as a prerequisite for a compensation decision to be made. 

SC&A notes that this review was triggered in part by concerns raised in a letter by Richard 

Miller (see Attachment 1) to the Secretary of HHS and the Director of NIOSH. Mr. Miller was 

concerned that the principal author of the Occupational Internal Dose TBD had serious conflict 

of interest (COI) issues and that the technical credibility of that TBD was questionable as well. 

As a result, particular emphasis has been placed on the review of that document. 

Construction of the PGDP began in 1951 and production started in 1952. The site is located in 

northeastern McCracken County, approximately 10 miles west of the City of Paducah, Kentucky, 

and covers 3,423 acres, of which 748 acres are within a security fence. Paducah was initially 

owned and operated by the U.S. Department of Energy (DOE) and its predecessor agencies. 

Since July 1, 1993, the DOE has leased the production facilities of the plant to the United States 

Enrichment Corporation (USEC). The USEC portion of the site is licensed and regulated by the 

Nuclear Regulatory Commission (USEC 1994). The DOE mission at the site is now one of 

Environmental Restoration and Waste Management Operations. In some cases, the DOE and 

USEC operations occur in the same building or area, resulting in potential exposure of USEC 

employees to DOE radioactive material and vice versa. 

Paducah has been actively involved in the enrichment of uranium, its primary function, since its 

inception. This has not been the sole activity of this site. In Building C-420, there was a facility 

to convert uranium trioxide (UO3) and urano-uranic oxide (U3O8) to uranium tetrafluoride (UF4) 

for further conversion in Building C-410. The C-410 Building was used for conversion of 

uranium tetrafluoride (UF4) to uranium hexafluoride (UF6), the feed material for the process 

buildings. Paducah was also involved in the reduction of UF6 to UF4, and subsequent conversion 

to metallic uranium. The process used was similar to that at the Fernald site. There are two 

maintenance shops for fabrication, assembly, maintenance, and repair of process equipment. The 

Cascade Improvement Program (CIP) in 1954–1961 and the Cascade Upgrade Program (CUP) in 

1973–1982 required isolating and opening portions of the cascade system to replace existing 

process equipment while the plant continued its uranium enrichment operations. These 

operations and the repair and upgrading of the equipment provided conditions for increased 
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radiation exposure of workers. An onsite smelter was used to salvage diffusion barriers and 

other sensitive items for PGDP and other government sites. Waste has either been stored above 

ground or taken to on-site burial areas (Donnelly 1984). PGDP participated in the work for other 

program, where they conducted projects for other sites. The DOE mission is currently 

Environmental Restoration and Waste Management Operations. 

SC&A’s review focused on the six TBDs that make up the PGDP site profile. These comprise 

Introduction, Site Description, Occupational Internal Dose, Occupational External Dose, 

Occupational Medical Dose, and Occupational Environmental Dose as these topics pertain to 

historic occupational radiation exposure of PGDP workers. These TBDs were issued in 2004– 

2006. As “living” documents, TBDs are constantly being revised as new information, 

experience, or issues arise. This review includes consideration of the August 2006 revision of 

the Site Description and the page change of the Introduction TBDs. The revised versions of the 

Occupational External Dose and Occupational Medical Dose, published in August and 

September of 2006, respectively, were not used for this review. SC&A’s process included a 

review of the TBDs, an onsite visit to conduct interviews with site experts and to identify 

documents for data retrieval, and reviews of retrieved PGDP and other historic records. The 

TBDs were evaluated for their completeness, technical accuracy, adequacy of data, compliance 

with stated objectives, and consistency with other site profiles, as stipulated in the SC&A 

Standard Operating Procedure for Performing Site Profile Reviews (SC&A 2004). A complete 

list of the PGDP TBDs and other supporting documents that were reviewed by SC&A is 

provided in Attachment 2. 

Interviews at the site were originally designed to solicit unclassified information; however, a 

number of site experts were more comfortable with classified interviews. Records reviews 

included both classified and unclassified documentation. As a result, it was also necessary to 

have interview notes and documents screened by PGDP classification reviewers to assure that no 

sensitive material was inadvertently included. Site expert interviews will be released after the 

interview summary has undergone appropriate classification review. At this time, additional 

information may be included in other sections of the report. 

Issues presented in this report are sorted into the following categories, in accordance with 

SC&A’s review procedures: 

(1) Completeness of data sources 

(2) Technical accuracy 

(3) Adequacy of data 

(4) Consistency among site profiles 

(5) Regulatory compliance 

Following the introduction and a description of the criteria and methods employed to perform the 

review, the report discusses the strengths of the TBD, followed by a description of the major 

issues identified during our review. The issues were carefully reviewed with respect to the five 

review criteria. Several of the issues were designated as primary findings, because they 

represent key deficiencies in the TBDs that need to be corrected and which have the potential to 

substantially impact at least some dose reconstructions. Others have been designated 

“observations” to both denote their importance for the technical adequacy and completeness of 
 

 

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the site profile, and to indicate that they have been judged by SC&A to have relatively less 

influence on dose reconstruction or the ultimate significance of worker doses so estimated. 

1.1 SUMMARY OF STRENGTHS 

In general, the Site Description TBD (Maisler 2006) provides a basic characterization of much of 

the history and operational activities at the PGDP. The major functions of many of the 

production and support buildings are outlined, as well as major process changes that occurred 

through the years. 

The authors of the TBDs benefited from having access to information and data that were 

compiled as a part of the PGDP programs, as follows: 

(1) Radiological control personnel have implemented improved procedures and technologies 

over time to reduce radiation dose to workers, and have improved personnel monitoring 

programs. 

(2) PGDP implemented an environmental monitoring program, including stack monitoring, 

perimeter monitoring, offsite monitoring, and groundwater monitoring. 

(3) Starting in 1961, all employees have been provided with an external dosimeter, creating a 

potentially useful coworker database. 

The PGDP TBDs provided a good description of the in-vivo and in-vitro monitoring programs 

over the period of operation, including detailed descriptions of dosimeters used for both 

beta/gamma and neutron monitoring. 

1.2 SUMMARY OF FINDINGS 

Finding 1: Uranium Enrichment Levels Achieved at PGDP could be Higher than 2% 

The Site Description TBD (Turpin 2006, pg. 7) states that PGDP enriched feed material (UF6) up 

to about 2.5% 235U. However, the default enrichment level specified in the Occupational Internal 

Dose TBD (Berger 2004) is only 2%. Although the assayed specific activity of 235U in the 

cascade product is consistent with this enrichment level, the maximum assayed specific activity 

of 234U is over 50% higher than the default activity of this isotope The higher activities provide 

the potential for higher radiation doses from 234U and perhaps from 235U than those calculated 

using the methodology prescribed in the TBD. 

Finding 2: The Number of Workers Assigned a Zero Dose should be Disclosed When 

Reporting the Mean of the Distribution of Doses 

The average recorded doses in Tables 2-2 and 2-3 of the Site Description TBD (Turpin 2006, 

pg. 11) are biased low and mean little without knowing the numbers of workers assigned dose 

values of zero when their measured dose was less than the minimum detectable level (MDL). 
 

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Finding 3: The TBD Needs to Consider Operations Other than Gaseous Diffusion 

There is no mention in the Site Description TBD (Turpin 2006) of the smelting operations that 

took place in Building C-746B or the smelting of diffusion barriers (including K-25 and 

Portsmouth barriers) during the cascade improvement and upgrade programs. These activities 

need to be described, since they could have resulted in significant doses to some workers. 

Paducah participated in the “Work for Others” (WFO) program providing assistance to other 

DOE sites. There is no mention of this program in the TBD and how this may affect the dose 

reconstructions in terms of radionuclides. 

Finding 4: The TBD Fails to Adequately Define and Assess Occupational Medical 

Exposure 

The guidelines, as presented by Kathren (2003), go a long way in assuring that all occupational 

medical exposures are reasonably included in determining the overall dose estimations for 

claimants. Unfortunately, these guidelines have not been applied in the TBD in a claimantfavorable manner. 

Finding 5: Contamination Control and Skin and Extremity Dose 

The TBDs do not provide enough information regarding the radiological controls in place (or 

lack thereof) for the operations at PGDP that pose the potential for exposures. Especially in the 

early days, lack of adequate radiological controls could lead to worker doses that may be missed 

in the process of dose reconstruction, particularly for unmonitored workers. Such workers may 

not be identifiable without this understanding of the effectiveness of the radiological controls in 

place and for work practices where significant dose was possible. Contamination control was a 

significant problem and should be examined for its relevance to skin dose. This should be 

considered in the context of dose reconstruction. Some method for assignment of extremity dose 

should be developed where this area is affected by cancer. 

Finding 6: Onsite Environmental Exposures are based on Site Boundary Data 

The TBD purports to describe “potential exposures from ambient sources while working outside 

the process buildings,” but the ambient monitoring data in the document are for measurements at 

the site boundaries and beyond. There are no corroborating data provided to demonstrate that 

these measurements are consistent with the levels that workers might experience while working 

at the site outside the process buildings. Two additional sources of ambient exposure include the 

burning of contaminated material in onsite pits (i.e., routine and incidentally), and the alleged 

intentional releases that occurred in the 1950s. 

Finding 7: Inadequate Characterization of the Source Term for Internal Exposures 

A critical issue is the specification of the source term for the internal exposures. The Internal 

Occupational Dose TBD (Berger 2004) does not fully utilize information presented in two key 

documents. One is the exposure assessment of workers at the PGDP (PACE/Utah 2000), which 

emphasized exposures to neptunium and plutonium. The other is the draft report on recycled 
 

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uranium mass balance (BJC 2001), which is cited as a source of information by Turpin (2006), 

but is not mentioned by Berger. 

Finding 8: Isotopic Fractions for Various Enrichments are not Properly Characterized in 

the TBD 

An examination of the data in Table 5-2 of the TBD reveals a number of discrepancies. The 

most glaring one is the specific activity of 235U in 93% enriched feed. A direct calculation yields 

an activity of 7.44E-2 Bq/μg, a factor of 10 higher than the listed value of 7.38E-3 Bq/μg. 

Finding 9: The Default Isotopic Distribution in Table 5-4 Ignores Many Isotopes 

Associated with Recycled Uranium 

Table 5-4 of the TBD lists default specific activities to be used when only total uranium results 

are available. This guidance ignores the information presented by Bechtel-Jacobs Corporation 

(BJC 2001, Table 2.4-1), which lists maximum concentrations of 99Tc, 237Np, and plutonium for 

11 different operations at various facilities in the PGDP in various time periods. NIOSH should 

evaluate these data for use in the dose reconstructions of workers involved in these operations. 

Finding 10: Particle Size of Inhaled Aerosols Assumed in the TBD is not Entirely Claimant 

Favorable 

In Table 5-5 of the TBD, the particle size for inhaled aerosols is specified as 5 μm Aerodynamic 

Medial Activity Diameter (AMAD). Such a particle-size assignment is not supported by the 

data, nor is it claimant favorable, because particle sizes significantly less than 5 μm AMAD are 

cited in the supporting literature. 

Finding 11: The List and Quantities of Transuranics Addressed in the TBD are not 

Complete or Claimant Favorable 

Table 5-5 of the TBD limits transuranics (TRU) to 237Np and 239Pu. However, the TRU in the 

Hanford reactor tails and other sources include 238Pu, 240Pu, and 241Am. In addition, information 

about the occurrence of TRU nuclides at different facilities is presented by PACE/Utah (2000, 

Appendix D), which lists various radiological data that need to be taken into consideration in the 

TBD. 

Finding 12: Lung Clearance Types need to be clearly Defined and Claimant Favorable 

The Lung Clearance Types, referred to as “Absorption Type” in Table 5-5, should be consistent 

with the chemical forms of each radionuclide. In case of uncertainty, the most claimantfavorable assumption should be adopted. The TBD is not always clear which chemical form 

and/or clearance type the dose reconstructor should assign to each element. 
 

 

 

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Finding 13: Radionuclide Intakes based on Bioassay Data Need to Take into Consideration 

Frequency of Sample Collection 

Table 5-6 of the TBD lists the frequency of in-vitro measurements at various facilities at the 

PGDP during various time periods. The frequencies range from once every 4 weeks to once a 

year. The last row of the table lists a default frequency of once every 4 weeks. Since the interval 

between measurements could have been as long as 1 year, this default assumption is not claimant 

favorable. The longer the interval over which the intakes occurred, be they chronic or acute, the 

lower the urine concentration for a given total intake. Conversely, the longer the elapsed time 

(for acute intakes) or exposure duration (for chronic) for a given urine concentration, the greater 

the derived intake. 

Finding 14: Minimum Detectable Concentrations are not Clearly Defined 

In many cases, we could not verify the listed MDCs. For example, the MDC for total uranium 

by in-house fluorimetry is listed as 10 μg/L. None of documents cited as data sources for this 

table list that value; both PACE/Utah (2000) and SAIC (1999) cite typical MDCs of 5 μg/L. The 

MDC for natural uranium assayed at ORNL from 1999 to the present is in error; the source 

document (SAIC 1999) lists the MDC at 0.06 μg/sample, while Table 5-7 lists it as 0.06 mg 

(60 μg), a 1000-fold discrepancy. The latter value is clearly inconsistent with the value of 5 μg/L 

for kinetic phosphorescence analysis (KPA), used in 1977–1982. Table 5-7 lists a default 

detection level of 0.27 pCi/L for urinalysis of individual isotopes of actinide elements (Th, U, 

Pu, and Am). The TBD cites ICRP Publication 54 (Annals of the ICRP Vol. 19 No. 1–3) , 

Individual Monitoring for Intakes of Radionuclides by Workers: Design and Interpretation 

(Oxford: Pergamon Press) as the source of these data, which are based on alpha spectroscopy.1 

However, Table 5-7 lists higher levels—1 pCi/L for Pu isotopes and 0.3 pCi/L for uranium 

isotopes. If the dose reconstructor does not know the analytical method involved, this default 

assumption is neither claimant favorable nor scientifically correct. 

Finding 15: Day of Sample Collection needs to be Taken into Consideration When 

Deriving Intakes based on Urine Analyses 

Section 5.3.3 of the TBD states the following: 

The practice of offsite collection of samples that takes place 24 to 48 hr after 

leaving the plant not only minimizes the possibility of sample crosscontamination, but it ensures that samples are collected after the transfer of the 

rapid clearance component. Some PGDP employees were asked to collect 

samples after 1 or 2 days off from work; if so, that collection instruction was 

sometimes noted on the analytical record. 

The TBD fails to note that this practice would lead to a lowering of the calculated intakes, nor 

does it instruct dose reconstructors to be alert to any cases for which urine samples were 

1 We also note that ICRP Publication 54 has been replaced by ICRP Publication 78: Individual Monitoring for 

Internal Exposure of Workers - Annals of the ICRP Volume 27/3-4, Replacement of ICRP Publication 54, 1998 

(Oxford:Pergamon Press). 
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collected after the worker was off work for any period of time. Appropriate adjustments to the 

calculated intake should be made to compensate for the lowered concentration in the urine 

following an absence from work. 

Finding 16: Additional Significant Incidents with Internal Dose Potential Need to be 

Discussed 

Significant information that could be useful to dose reconstructors is not included in the TBD. 

This information includes statements by workers that urine specimens were collected within 

30 minutes of an incident or accident, with a potential for elevated exposure. Such a short time 

period does not allow for equilibrium between the inhaled activity and the concentration in the 

urine, most likely resulting in a false negative. Follow-up samples were collected from workers 

who did show elevated levels of radioactive materials in the urine. More important are the 

statements of former workers that the bioassays performed following such incidents were not 

always recorded in the database. Thus, the doses from such incidents may not be in the worker 

exposure records. 

Finding 17: The Coworker Model for Applying Bioassay Data to Unmonitored Workers is 

neither Scientifically Valid nor Claimant Favorable 

In the coworker models used by NIOSH, workers are not classified by their jobs or by the 

buildings where they performed their work. In the TIB describing the coworker model for 

internal dosimetry (Ikenberry 2005), there is no attempt to sort the urinalysis results by job 

assignment or location to determine if there were any correlations between the uranium 

concentration in the urine and the building or department where the job was performed. 

In order to apply the coworker model to unmonitored workers, NIOSH needs to demonstrate that 

there is a low probability that any unmonitored worker could have higher exposures than the 

monitored workers taken as a group. Ikenberry (2005) fails to do so. 

Finding 18: The Method of Converting Mass Concentrations of Uranium in Urine to 

24-hour Excretions of Activity of Uranium Isotopes is neither Scientifically Valid nor 

Claimant Favorable 

The TIB describing the coworker model for internal dosimetry (Ikenberry 2005) uses the nonclaimant-favorable assumption about the specific activities of uranium isotopes presented in the 

internal dose TBD (Berger 2004). The default specific activity should be increased from 

0.0389 Bq/μg to 0.0541 Bq/μg. The daily excretion of urine should be updated to reflect the 

latest ICRP (2002) recommendations: 1.6 L/d for male workers and 1.2 L/d for females. 

Finding 19: Shallow Dose from Beta Emitters may have been Underestimated 

According to the Occupational External Dose TBD (Turner 2005), the film badges used to derive 

skin doses from beta emitters employed a minimum absorber thickness of 80 mg/cm2

 between 

the film and the source, but the film badges appear to have been calibrated with a uranium slab 
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without the absorber. Under these conditions, a large portion of the skin dose from weak betas 

emitted by uranium would not be detected, and the skin dose significantly underestimated. 

Finding 20: Questionable Assumptions for Assigning Skin and Deep Dose for 

Unmonitored Workers Prior to 1960 by means of Coworker Data 

SC&A is concerned that the coworker model described in the Occupational External Dose TBD 

(Turner 2005) is based on the assumption that prior to 1960, the population of “. . . monitored 

individuals represents those with the highest exposure potential.” It is SC&A’s contention that 

these monitoring data reflect a badging practice that not only included all worker categories 

(regardless of their potential for exposure), but furthermore diluted the average dose within a 

given worker category by rotating badge assignments. However, this issue appears to have been 

resolved with the issuance of ORAUT-OTIB-0031 (Merwin 2005). 

Finding 21: Assessment of Neutron Exposures at PGDP Appears to have been 

Underestimated 

The principal sources of neutron exposures at PGDP involve the α, n reaction with fluorine 

compounds (UF4, UF6) and exposures to radionuclides that undergo spontaneous fission (i.e., 

TRU). However, based on the Occupational External Dose TBD (Turner 2005), it appears that 

reliable monitoring of neutron exposures did not begin until 1998, and NIOSH had to account for 

missed neutron dose based on worker activities and models that employ neutron-to-photon ratios. 

SC&A is concerned that the coworker model may not be dependable for exposures prior to 1961, 

due to a paucity of photon dosimetry at that time. 

Another document is the draft report on recycled uranium mass balance (BJC 2001), which is 

cited as a source of information by Turpin (2006), but is not mentioned in this TBD. That report 

breaks down the potential exposures to three contaminants in recycled uranium—99Tc, 237Np, 

and plutonium—into 11 types of activities and operations at the PGDP, and lists the maximum 

concentrations of each of these constituents (BJC 2001, Table 2.4-1). 
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2.0 SCOPE AND INTRODUCTION 

This review of the Paducah Gaseous Diffusion Plant (PGDP) site profile was conducted from 

April 1 through August 31, 2006, in support of the Advisory Board on Radiation and Worker 

Health (Advisory Board) in the latter’s statutory responsibility under the Energy Employees 

Occupational Illness Compensation Program Act of 2000 (EEOICPA) to conduct such reviews 

and advise the Secretary of Health and Human Services (HHS) on the “completeness and 

adequacy” of the EEOICPA program. This review included interviews of site experts by S. 

Cohen & Associates (SC&A) regarding specific Technical Basis Documents (TBDs) that make 

up the PGDP site profile. The review was limited to unclassified materials generated from 

worker interviews, and document reviews were submitted for classification screening to ensure 

that no sensitive information was inadvertently included in this report. 

The review was triggered, in part, by concerns raised in a letter by Richard Miller (see 

Attachment 1) to the Secretary of HHS and the Director of NIOSH. Miller was concerned that 

the principal author of the internal dose TBD had serious COI issues, and that the technical 

credibility of that TBD was questionable as well. As a result, particular emphasis has been 

placed on the internal dose portions of the site profile. 

A subgroup of PGDP workers has been designated as an SEC, as defined in 42 CFR Part 83: 

The Special Exposure Cohort (‘‘the Cohort’’) is a category of employees defined 

under 42 U.S.C. 7384l(14). In this definition, Congress specified classes of 

employees to comprise the Cohort initially, including DOE employees, DOE 

contractor or subcontractor employees, who were (1) employed an aggregate of 

at least 250 work days before February 1, 1992 at a gaseous diffusion plant in 

Paducah, Kentucky, Portsmouth, Ohio, or Oak Ridge, Tennessee, and (2) who 

were monitored using dosimetry badges or worked in a job that had exposures 

comparable to a job that is or was monitored using dosimetry badges.... 

As provided in 42 U.S.C. 7384l(9)(A), employees included in the Cohort who 

incur a specified cancer qualify for compensation (see DOL regulations 20 CFR 

part 30 for details). Cancer claims submitted by these employees or their 

survivors do not require DOL to evaluate the probability that the cancer was 

caused by radiation doses incurred during the performance of duty for nuclear 

weapons programs of DOE, as is required for other cancer claims covered by 

EEOICPA. 

Specified cancers are a limited group of cancers that EEOICPA specifies are 

compensable under provisions governing compensation for members of the 

Cohort. Although the list of specified cancers is determined by statute, the list 

can also be found in this rule under § 83.5. 

83.5(m) Specified cancer, as is defined in Section 3621(17) of EEOICPA (42 

U.S.C. 7384l(17)) and the DOL regulation implementing EEOICPA (20 CFR 

30.5(dd)), means: 
 

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(1) Leukemia (other than chronic lymphocytic leukemia) provided that onset 

of the disease was at least two years after initial occupational exposure; 

(2) Lung cancer (other than in situ lung cancer that is discovered during or 

after a post-mortem exam); 

(3) Bone cancer; 

(4) Renal cancers; 

(5) The following diseases, provided onset was at least 5 years after first 

exposure: 

(i) Multiple myeloma; 

(ii) Lymphomas (other than Hodgkin’s disease); 

(iii) Primary cancer of the: 

(A) Thyroid; 

(B) Male or female breast; 

(C) Esophagus; 

(D) Stomach; 

(E) Pharynx; 

(F) Small intestine; 

(G) Pancreas; 

(H) Bile ducts; 

(I) Gall bladder; 

(J) Salivary gland; 

(K) Urinary bladder; 

(L) Brain; 

(M) Colon; 

(N) Ovary; 

(O) Liver (except if cirrhosis or hepatitis B is indicated). 

(6) The specified diseases designated in this section mean the physiological 

condition or conditions that are recognized by the National Cancer 

Institute under those names or nomenclature, or under any previously 

accepted or commonly used names or nomenclature. 

While the SEC designation allows a subgroup of workers to file claims under the Act without 

requiring a dose reconstruction, other workers with uncovered cancers may require a dose 

reconstruction to be eligible for compensation. The integrity of the dose reconstruction process 

remains a critical consideration. 

2.1 REVIEW SCOPE 

Under the EEOICPA and Federal regulations defined in 42 CFR Part 82, Methods for Radiation 

Dose Reconstruction Under the Energy Employees Occupational Illness Compensation 

Program, of the Code of Federal Regulations, the Advisory Board is mandated to conduct an 

independent review of the methods and procedures used by NIOSH and its contractors for dose 

reconstruction. As a contractor to the Advisory Board, SC&A has been charged under Task 

Order 1 to support the Advisory Board in this effort by independently evaluating a select number 

of site profiles that correspond to specific facilities at which energy employees worked and were 

exposed to ionizing radiation. 
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This report provides a review of the following six documents related to historical occupational 

exposures at the PGDP site: 

• 	 Maisler, Daniel 2006, Technical Basis Document for Paducah Gaseous Diffusion Plant – 

Introduction,), ORAUT-TKBS-0019-1, (Rev. 00 PC-1), ORAU Team Dose 

Reconstruction Project for NIOSH, August 8, 2006. 

• 	 Turpin, Baynard, 2006, Technical Basis Document for the Paducah Gaseous Diffusion 

Plant – Site Description, ORAUT-TKBS-0019-2, (Rev. 01), ORAU Team Dose 

Reconstruction Project for NIOSH, August 21, 2006. 

• 	 Turner, James E., 2004, Technical Basis Document for the Paducah Gaseous Diffusion 

Plant – Occupational Medical Dose, ORAUT-TKBS-0019-3, (Rev. 00), ORAU Team 

Dose Reconstruction Project for NIOSH, February 11, 2004. 

• 	 East, James, 2004, Technical Basis Document for the Paducah Gaseous Diffusion Plant – 

Occupational Environmental Dose, ORAUT-TKBS-0019-4, (Rev. 00), ORAU Team Dose 

Reconstruction Project for NIOSH, August 24, 2004. 

• 	 Berger, Carol D., 2004, Technical Basis Document for Paducah Gaseous Diffusion Plant 

– Occupational Internal Dose, ORAUT-TKBS-0019-5, (Rev. 00), ORAU Team Dose 

Reconstruction Project for NIOSH, September 30, 2004. 

• 	 Turner, James E, 2005, Paducah Gaseous Diffusion Plant – Occupational External Dose, 

ORAUT-TKBS-0019-6, (Rev. 01), ORAU Team Dose Reconstruction Project for NIOSH, 

March 29, 2005. 

These documents are supplemented by two technical information bulletins (TIBs), which provide 

additional guidance to the dose reconstructor: 

• 	 Ikenberry, Tracy A., 2005, Internal Dosimetry Coworker Data for Paducah Gaseous 

Diffusion Plant, ORAUT-OTIB-0037, (Rev 00), ORAU Team Dose Reconstruction 

Project for NIOSH, September 20, 2005 

• 	 Merwin, Steven E., 2005, External Coworker Dosimetry Data for the Paducah Gaseous 

Diffusion Plant, ORAUT-OTIB-0031, (Rev 00 PC-1), ORAU Team Dose Reconstruction 

Project for NIOSH, August 15, 2005. 

Since these TIBs supplement specific TBDs in the site profile, they were also considered in this 

review. 

Implementation guidance is also provided by so-called “workbooks,” which have been 

developed by NIOSH for selected sites to provide more definitive direction to the dose 

reconstructors on how to interpret and apply TBDs, as well as other available information. The 

SC&A team did not include review of workbooks related to the PGDP site profile (i.e., the 
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workbook accompanying ORAUT-OTIB-0031). This workbook is being reviewed under 

another task order. 

SC&A, in support of the Advisory Board, has critically evaluated the PGDP Site TBDs for the 

following: 

• 	 Determine the completeness of the information gathered by NIOSH in behalf of the site 

profile, with a view to assessing its adequacy and accuracy in supporting individual dose 

reconstructions 

• 	 Assess the technical merit of the data/information 

• 	 Assess NIOSH’s use of the data in dose reconstructions 

SC&A’s review of the six TBDs focuses on the quality and completeness of the data that 

characterizes the facility and its operations, and the use of these data in dose reconstruction. The 

review was conducted in accordance with SC&A Standard Operating Procedure for Performing 

Site Profile Reviews (SC&A 2004), which was approved by the Advisory Board. 

The review is directed at “sampling” the site profile analyses and data for validation purposes. 

The review does not provide a rigorous quality control process, whereby actual analyses and 

calculations are duplicated or verified. The scope and depth of the review are focused on aspects 

or parameters of the site profile that would be particularly influential in deriving dose 

reconstructions, bridging uncertainties, or correcting technical inaccuracies. This review does 

not explicitly address the issue of radiation exposures to cleanup workers and decommissioning 

workers, as that is not addressed in the TBDs. 

The six TBDs serve as site-specific guidance documents used in support of dose reconstructions. 

These site profiles provide the health physicists who conduct dose reconstructions on behalf of 

NIOSH with consistent general information and specifications to support their individual dose 

reconstructions. This report was prepared by SC&A to provide the Advisory Board with an 

evaluation of whether and how the TBDs can support dose reconstruction decisions. The criteria 

for evaluation include whether the TBDs provide a basis for scientifically supportable dose 

reconstruction in a manner that is adequate, complete, efficient, and claimant favorable. 

Specifically, these criteria were viewed from the lens of whether dose reconstructions based on 

the TBDs would provide for robust compensation decisions. 

The basic principle of dose reconstruction is to characterize the radiation environments to which 

workers were exposed and determine the level of exposure the worker received in that 

environment through time. The hierarchy of data used for developing dose reconstruction 

methodologies is dosimeter readings and bioassay data, coworker data and workplace monitoring 

data, and process description information or source term data. 

2.2 REVIEW APPROACH 

SC&A’s review of the TBDs and supporting documentation concentrated on determining the 

completeness of data collected by NIOSH, the adequacy of existing PGDP personnel and 
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environmental monitoring data, and the evaluation of key dose reconstruction assumptions. Site 

expert interviews were conducted at the PGDP, and all notes taken were submitted for screening 

by appropriate classifiers. 

Site expert interviews were conducted to help SC&A obtain a comprehensive understanding of 

the radiation protection program, site operations, and environmental contamination. Interviews 

were conducted by SC&A by teleconference or in person during the course of this review. The 

site experts included current and former staff from radiation control, operations, environmental 

monitoring, maintenance, and other organizations, as well as other site experts knowledgeable of 

particular elements of the PGDP environmental safety and health program. These individuals 

were given the opportunity to review the interview summary for accuracy. This is an important 

safeguard against missing key issues or misinterpreting some vital piece of information. A 

master summary of all interviews conducted is reserved pending review for classification 

purposes, and will be provided at a later date. 

2.3 REPORT ORGANIZATION 

In accordance with directions provided by the Advisory Board and with site profile review 

procedures prepared by SC&A and approved by the Advisory Board, this report is organized into 

the following sections: 

(1) Executive Summary 

(2) Scope and Introduction 

(3) Assessment Criteria and Method 

(4) Site Profile Strengths 

(5) Vertical Issues 

(6) Overall Adequacy of the Site Profile as a Basis for Dose Reconstruction 

Based on the issues raised in each of these sections, SC&A prepared a list of findings, which are 

provided in the executive summary. Issues are designated as findings if SC&A believes that they 

represent deficiencies in the TBD that need to be corrected, and which have the potential to have 

a substantial impact on at least some dose reconstructions. Issues can also be designated as 

observations if they simply raise questions, which, if addressed, would further improve the TBDs 

and may possibly reveal deficiencies that will need to be addressed in future revisions of the 

TBDs. 

Many of the issues that surfaced in the report correspond to more than one of the major 

objectives (i.e., strengths, completeness of data, technical accuracy, consistency among site 

profiles, and regulatory compliance). Section 6.0 provides a list of the issues in summary form, 

and to which objective each particular issue applies. 

In many ways, the TBDs have done a successful job in addressing a series of technical 

challenges. In other areas, the TBDs exhibit shortcomings that may influence some dose 

reconstructions in a substantial manner. 
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3.0 ASSESSMENT CRITERIA AND METHODS 

SC&A is charged with evaluating the approach set forth in the site profiles that is used in the 

individual dose reconstruction process. These documents are reviewed for their completeness, 

technical accuracy, adequacy of data, consistency with other site profiles, and compliance with 

the stated objectives, as defined in SC&A Standard Operating Procedure for Performing Site 

Profile Reviews (SC&A 2004). This review is specific to the PGDP Site Profile and supporting 

TIBs; however, items identified in this report may be applied to other facilities, especially 

facilities with similar source terms and exposure conditions. The review identifies a number of 

issues and discusses the degree to which the site profile fulfills the review objectives delineated 

in SC&A’s site profile review procedure. 

3.1 OBJECTIVES 

SC&A reviewed the site profile with respect to the degree to which technically sound judgments 

or assumptions are employed. In addition, the review identifies assumptions by NIOSH that give 

the benefit of the doubt to the claimant. 

3.1.1 Objective 1: Completeness of Data Sources 

SC&A reviewed the site profile with respect to Objective 1, which requires SC&A to identify 

principal sources of data and information that are applicable to the development of the site 

profile. The two elements examined under this objective include (1) determining if the site 

profile made use of available data considered relevant and significant to the dose reconstruction, 

and (2) investigating whether other relevant/significant sources are available, but were not used 

in the development of the site profile. For example, if data are available in site technical reports 

or other available site documents for particular processes, and if the TBDs have not taken into 

consideration these data where it should have, this would constitute a completeness of data issue. 

The Oak Ridge Associated Universities (ORAU) site profile document database, including the 

referenced sources in the TBDs, was evaluated to determine the relevance of the data collected 

by NIOSH to the development of the site profile. Additionally, SC&A evaluated records 

supplied by USEC and publicly available records relating to the PGDP site and records provided 

by site experts. 

3.1.2 Objective 2: Technical Accuracy 

SC&A reviewed the site profile with respect to Objective 2, which requires SC&A to perform a 

critical assessment of the methods used in the site profile to develop technically defensible 

guidance or instruction, including evaluating field characterization data, source term data, 

technical reports, standards and guidance documents, and literature related to processes that 

occurred at PGDP. The goal of this objective is to first analyze the data according to sound 

scientific principles, and then to evaluate this information in the context of compensation. For 

example, if SC&A found that the technical approach used by NIOSH was not scientifically 

sound or claimant favorable, this would constitute a technical accuracy issue. 
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3.1.3 Objective 3: Adequacy of Data 

SC&A reviewed the site profile with respect to Objective 3, which requires SC&A to determine 

whether the data and guidance presented in the site profile are sufficiently detailed and complete 

to conduct dose reconstruction, and whether a defensible approach has been developed in the 

absence of data. In addition, this objective requires SC&A to assess the credibility of the data 

used for dose reconstruction. The adequacy of the data identifies gaps in the facility data that 

may influence the outcome of the dose reconstruction process. For example, if a site did not 

monitor all workers exposed to neutrons who should have been monitored, this would be 

considered a gap and, therefore, an inadequacy in the data. 

3.1.4 Objective 4: Consistency Among Site Profiles 

SC&A reviewed the site profile with respect to Objective 4, which requires SC&A to identify 

common elements within site profiles completed or reviewed to date, as appropriate. This 

objective was accomplished by reviewing key TBD assumptions for determining medical, 

environmental, internal and external dose from PGDP, and previously reviewed TBDs. This 

assessment was conducted to identify areas of inconsistencies, and determine the potential 

significance of any inconsistencies with regard to the dose reconstruction process. 

3.1.5 Objective 5: Regulatory Compliance 

SC&A reviewed the site profile with respect to Objective 5, which requires SC&A to evaluate 

the degree to which the site profile complies with stated policy and directives contained in 

42 CFR Part 82. In addition, SC&A evaluated the TBD for adherence to general quality 

assurance policies and procedures utilized for the performance of dose reconstructions. 

In order to place the above objectives into the proper context as they pertain to the site profile, it 

is important to briefly review key elements of the dose reconstruction process, as specified in 

42 CFR Part 82. Federal regulations specify that a dose reconstruction can be broadly placed 

into one of three discrete categories. These three categories differ greatly in terms of their 

dependence on and the completeness of available dose data, as well as on the 

accuracy/uncertainty of data. 

Category 1: Least challenged by any deficiencies in available dose/monitoring data are dose 

reconstructions for which even a partial assessment (or minimized dose(s)) corresponds to a 

probability of causation (POC) value in excess of 50%, and assures compensability to the 

claimant. Such partial/incomplete dose reconstructions with a POC greater than 50% may, in 

some cases, involve only a limited amount of external or internal data. In extreme cases, even a 

total absence of a positive measurement may suffice for an assigned organ dose that results in a 

POC greater than 50%. For this reason, dose reconstructions in behalf of this category may only 

be marginally affected by incomplete/missing data or uncertainty of the measurements. In fact, 

regulatory guidelines recommend the use of a partial/incomplete dose reconstruction, the 

minimization of dose, and the exclusion of uncertainty for reasons of process efficiency, as long 

as this limited effort produces a POC of greater than or equal to 50%. 
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Category 2: A second category of dose reconstruction is defined by Federal guidance, which 

recommends the use of “worst-case” assumptions. The purpose of worst-case assumptions in 

dose reconstruction is to derive maximal or highly improbable dose assignments. For example, a 

worst-case assumption may place a worker at a given work location 24 hours per day and 

365 days per year. The use of such maximized (or upper bound) values, however, is limited to 

those instances where the resultant maximized doses yield POC values below 50%, which are 

not compensated. For this second category, the dose reconstructor needs only to ensure that all 

potential internal and external exposure pathways have been considered. 

The obvious benefit of worst-case assumptions and the use of maximized doses in dose 

reconstruction is efficiency. Efficiency is achieved by the fact that maximized doses avoid the 

need for precise data and eliminates consideration for the uncertainty of the dose. Lastly, the use 

of bounding values in dose reconstruction minimizes any controversy regarding the decision not 

to compensate a claim. 

Although simplistic in design, to satisfy this type of a dose reconstruction, the TBD must, at a 

minimum, provide information and data that clearly identify (1) all potential radionuclides, 

(2) all potential modes of exposure, and (3) upper limits for each contaminant and mode of 

exposure. Thus, for external exposures, maximum dose rates must be identified in time and 

space that correspond to a worker’s employment period, work locations, and job assignment; 

similarly, in order to maximize internal exposures, highest air concentrations and surface 

contaminations must be identified. 

Category 3: The most complex and challenging dose reconstructions consist of claims where 

the case cannot be dealt with under one of the two categories above. For instance, when a 

minimum dose estimate does not result in compensation, a next step is required to make a more 

complete estimate. Or when a worst-case dose estimate that has assumptions that may be 

physically implausible results in a POC greater than 50%, a more refined analysis is required. A 

more refined estimate may be required either to deny or to compensate. In such dose 

reconstructions, which may be represented as “reasonable,” NIOSH has committed to resolve 

uncertainties in favor of the claimant. According to 42 CFR Part 82, NIOSH interprets 

“reasonable estimates” of radiation dose to mean the following: 

. . . estimates calculated using a substantial basis of fact and the application of 

science-based, logical assumptions to supplement or interpret the factual basis. 

Claimants will in no case be harmed by any level of uncertainty involved in 

their claims, since assumptions applied by NIOSH will consistently give the 

benefit of the doubt to claimants. [Emphasis added.] 

In order to achieve the five objectives described above, SC&A reviewed each of the six TBDs, 

and TIBs, giving due consideration to the three categories of dose reconstructions that the site 

profile is intended to support. The six TBDs provide reasonably well-organized information for 

the dose reconstructor when adequate data were available to do that comprehensively, as 

described below: 
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Technical Basis Document for Paducah Gaseous Diffusion Plant – Introduction, ORAUTTKBS-0019-1, Rev. 00 PC-1, (Maisler 2006), explains the purpose and the scope of the TBDs 

comprising the site profile. SC&A was attentive to this section, because it explains the role of 

each TBD in support of the dose reconstruction process. SC&A recognizes that the site profile is 

not required by the EEOICPA or by 42 CFR Part 82, which implements the statute. Site profiles 

were developed by NIOSH as a resource to the dose reconstructors for identifying site-specific 

practices, parameter values, and factors that are relevant to dose reconstruction. Based on 

information provided by NIOSH personnel, SC&A understands that site profiles are living 

documents, which are revised, refined, and supplemented with TIBs as required to help dose 

reconstructors. Site profiles are not intended to be prescriptive nor necessarily complete in terms 

of addressing every possible issue that may be relevant to a given dose reconstruction. Hence, 

the introduction helps in framing the scope of the site profile. NIOSH may want to include 

additional qualifying information in the introduction to this and other site profiles describing the 

dose reconstruction issues that are not explicitly addressed by a given site profile. 

Technical Basis Document for Paducah Gaseous Diffusion Plant – Site Description, ORAUTTKBS-0019-2, Rev. 01, (Turpin 2006), is an extremely important document, because it provides 

a description of the facilities, processes, and historical information that serve as the underpinning 

for subsequent PGDP TBDs. SC&A’s review of this section specifically addresses whether all 

of the potentially important site activities and processes are described, and whether 

characterization of source terms is complete and sufficient to support dose reconstruction. 

Technical Basis Document for Paducah Gaseous Diffusion Plant – Occupational Medical 

Dose, ORAUT-TKBS-0019-3, Rev. 00, (Turner 2004), provides an overview of the 

sources, types of exposure, and the frequency of exams that workers potentially received. 

The TBD clearly acknowledges the paucity of actual data to substantiate doses or support 

individual worker dose contributions—especially in the years prior to1975. The TBD 

draws heavily upon assessments at other sites, e.g., the Mound Plant and the Los Alamos 

National Laboratory (LANL). SC&A reviewed this section for technical adequacy and 

consistency with other NIOSH TBDs and procedures. 

Technical Basis Document for Paducah Gaseous Diffusion Plant – Occupational Environmental 

Dose, ORAUT-TKBS-0019-4, Rev. 00, (East 2004), provides background information and 

guidance to dose reconstructors for reconstructing the doses to unmonitored workers outside the 

facilities at the site who may have been exposed to routine and episodic exposures from these 

facilities. SC&A reviewed this section from the perspective of the source terms, measurements, 

and assumptions used to derive the external and internal doses to these workers. 

Technical Basis Document for Paducah Gaseous Diffusion Plant – Occupational Internal Dose, 

ORAUT-TKBS-0019-5, Rev. 00, (Berger 2004), presents background information and guidance 

to dose reconstructors for deriving occupational internal doses to workers. This section was 

reviewed with respect to background information and guidance regarding the types, mixes, and 

chemical forms of the radionuclides that may have been inhaled by the workers, the 

recommended assumptions for use in reconstructing internal doses based on in-vivo and in-vitro 

data, the methods recommended for use in the reconstruction of missed internal dose, and the 

methods recommended for characterizing uncertainty in the reconstructed internal doses. 
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Technical Basis Document for Paducah Gaseous Diffusion Plant – Occupational External Dose, 

ORAUT-TKBS-0019-6, Rev. 01, (Turner 2005), presents background information and guidance 

to dose reconstructors for deriving occupational external doses to workers. This section was 

reviewed with respect to background information and guidance regarding the different types of 

external radiation (i.e., gamma, beta, and neutron) and the energy distribution of this radiation to 

which the workers may have been exposed. We also reviewed the recommendations for 

converting external dosimetry data to organ-specific doses, the methods recommended for use in 

the reconstruction of missed external doses, and the methods recommended for characterizing 

uncertainty in the reconstructed external doses. 

SC&A has not developed a list of questions, and has not held conference calls with NIOSH to 

help clarify our understanding of the TBDs. Hence, this review has not had the benefit of this 

step in the site profile review process. 

Comparisons have been made between the methodologies used in the PGDP site profile and 

other TBDs reviewed to date. In the case of occupational medical exposure, the limited data 

available warranted additional comparison to TBDs for other sites. These comparisons focus on 

the methodologies and assumptions associated with dose reconstruction and resultant values used 

to obtain a POC (see Attachment 4). 

There are three levels of review typically employed in the preparation of a site profile review. 

First, SC&A team members review the report internally (this review has been performed). 

Second, SC&A project management reviews all aspects of this report (which has been 

performed), and, if necessary, outside experts are called upon to review selected portions of this 

report. It was deemed not necessary to engage an independent outside expert review, due to the 

extensive internal review this report received. The third level, referred to as the expanded review 

cycle, will consist of a review of this draft report by an Advisory Board-designated Working 

Group. The issues matrix, which will be provided under separate cover, will be used as the 

starting point to discuss and resolve issues. 

As the Working Group addresses the various issues raised in this report, it will report to the 

Advisory Board on the status of issues resolution. At such time that the Advisory Board deems 

appropriate, they will direct SC&A to perform additional analysis and or prepare revisions to this 

draft report to reflect the findings of the Working Group and the Advisory Board. 

Finally, SC&A notes that its review of the six TBDs and their supporting TIBs is not exhaustive. 

These are large, complex documents and SC&A used its judgment in selecting those issues that 

we believe are important with respect to dose reconstruction. 
 

 

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4.0 SITE PROFILE STRENGTHS 

In developing a TBD, the assumptions used must be fair, consistent, and scientifically robust, 

and uncertainties and inadequacies in source data must be explicitly addressed. The 

development of the TBD must also consider efficiency in the process of analyzing individual 

exposure histories, so that claims can be processed in a timely manner. With this perspective in 

mind, we identified a number of strengths in the PGDP site TBDs. These strengths are described 

in the following sections. 

4.1 COMPLETENESS OF DATA 

In general, the Site Description TBD (Maisler 2006) provides a basic characterization of much of 

the history and operational activities at the PGDP. The major functions of many of the 

production and support buildings are outlined, as well as major process changes that occurred 

through the years. There are still some gaps related to processing of special materials and 

recycled uranium. The text provides a short list of special incidents and activities with a high 

potential for increased radiation exposure for the period 1956–1981. This information can assist 

dose reconstructors in determining whether additional exposure potential exists, and whether the 

total dose reconstruction process is claimant favorable. That said, the information presented is a 

brief summary of the available data. A more comprehensive treatment of these subjects can be 

found in the PACE/Utah (2000) and Bechtel (BJC 2001) reports. In any case, NIOSH will often 

have to reconstruct doses for workers in the period before about 1974—and especially before 

1961—from very limited film badge and bioassay data. In-vitro bioassay data are generally 

limited to measurements of the total mass of uranium in urine. Radionuclide-specific estimates 

of activity intakes must be based on supplementary tables of activity relative to the total mass of 

uranium from those measurements. 

4.2 ADEQUACY OF DATA 

The TBDs benefited from having access to information and data that were compiled as a part of 

the PGDP programs, as follows: 

(1) Radiological control personnel have implemented improved procedures and technologies 

over time to reduce radiation dose to workers, and have improved personnel monitoring 

programs. 

(2) PGDP implemented an environmental monitoring program, including stack monitoring, 

perimeter monitoring, offsite monitoring, and groundwater monitoring. 

(3) Starting in 1961, all employees have been provided with an external dosimeter. 

Although PGDP has significant quantities of personnel monitoring data, as well as 

environmental data, there are gaps in the information. Only a fraction of the population was 

monitored for radiation exposure in the early years of operation. 
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4.3 TECHNICAL ACCURACY/CLAIMANT FAVORABILITY 

The PGDP TBDs exhibited the following strengths in terms of their technical accuracy and 

claimant favorability: 

(1) NIOSH has provided a good description of the in-vivo and in-vitro monitoring programs 

over the period of operation. They have provided detailed descriptions of dosimeters 

used for both beta/gamma and neutron monitoring. The TBDs and OTIBs made efforts 

to track the changes in dosimetry methods, calibration standards, and administrative 

limits for the period of 1952 to 2003 to assist in making the assigned dose claimant 

favorable. Chronological sequences of events at the PGDP were outlined in the TBDs, 

along with associated tables of dosimetry methodology, sensitivity, and limitations as 

dosimetry changed over the years. Major areas of external radiation hazards were 

addressed as the functions of the PGDP changed over time. 

(2) The dose reconstructor generally uses the inhalation Absorption Type (F, M, or S) that 

gives the best outcome to the claimant. 
 

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5.0 VERTICAL ISSUES 

This section presents a list and brief description of the issues identified with each TBD. No 

distinction is made regarding the relative importance of the issues. However, the more important 

issues are designated as findings in the Executive Summary. 

5.1 SITE DESCRIPTION TBD 

5.1.1 Issue 1: The Enrichment Facilities at PGDP are no longer under DOE Regulation 

Although Section 2.1.1 of the Site Description TBD (Turpin 2006) states that “the purpose of the 

gaseous diffusion plant has been and continues to be the enrichment of uranium...”, there is no 

indication in the site profile TBDs that since July 1, 1993, the United States Enrichment 

Corporation (USEC) has performed these operations at the site under the terms of a lease from 

DOE, and that those parts of the plant used by USEC are currently operated in compliance with 

NIOSH and NRC, rather than DOE regulations. The remaining portions of the plant are still 

under DOE control and regulation. DOE is responsible for DOE Material Storage Areas. Much 

of this material is stored in otherwise unused portions of buildings leased by USEC. The data 

reported in the site profile TBDs since 1993 reflect only the USEC portion of the site operations. 

DOE is responsible for environmental restoration and decommissioning at PGDP, and has been 

since July 1, 1993. 

For some facilities, there may be both USEC and DOE activities, such as the storage of 

radioactive material in cascade buildings. This not only results in exposure to the DOE source 

term, but also the USEC source term. A methodology for addressing exposure to workers from 

both agencies should be clarified in the TBD. This is especially pertinent since the DOE and 

NRC implement different radiation protection regulations. 

USEC anticipates ending its lease of the plant from the DOE at some time in the future. Prior to 

USEC taking possession of leased areas at PGDP, a major site characterization study was 

performed. This established a baseline for site contamination, such that USEC would not be held 

responsible for site contamination the DOE had created. Although SC&A is working with DOE 

to obtain this report, USEC has not released it to date. This document would provide valuable 

information on the existing radiological conditions in the environment, as well as the workplace. 

5.1.2 Issue 2: The Enrichment Level at PGDP is no longer Limited to About 2.5% 

Section 2.1.1 of Turpin (2006) states that PGDP enriched natural uranium up to about 2.5% 235U. 

Footnote 1 on page 5 of the Occupational Internal Dose TBD (Berger 2004) indicates that the 

predominant level of the enriched product was 1.5%, although enrichments of up to 5% were 

eventually performed. The higher enrichment levels could provide the potential for increased 

radiation doses from 234U and 235U relative to those for 238U. 
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5.1.3 Issue 3: The Site Map in the TBD is Inadequate 

The site description in Turpin (2006) provides very little information in terms of layout of 

process buildings, office buildings, storage and waste areas, holding ponds, etc. Unfortunately, 

the site map (Figure 2-2) is also uninformative. The PACE/Utah 2000 report site map (see 

Figure 5.1-1) shows the relative locations of the principal buildings and also identifies those that 

are currently leased to USEC. A good site map could provide the dose reconstructor with 

information such as the location of a worker relative to a nearby exposure source that might 

otherwise be overlooked. 

Figure 5.1-1. Schematic Drawing Showing Approximate Locations of PGDP Buildings 

(Source: PACE/Utah 2000, Figure C-1) 

5.1.4 Issue 4: The Material Processed to make Feed Material was not Just UO3 

The Site Description TBD identifies the material used to produce feed material as UO3 (pp. 6 and 

17). U3O8, which also was used in producing feed material (pg. 18), was also processed. Since 

U3O8 is classified as Type S for inhalation, while UO3 is Type M, the distinction can be 

important in dose reconstruction. 
 

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5.1.5 Issue 5: Average External Doses are biased Low by the Use of Assigned Zero 

Values 

The average recorded doses in Tables 2-1, 2-2 and 2-3 of Turpin 2006 are biased low and mean 

little without knowing the numbers of workers assigned dose values of zero when their measured 

dose was less than the MDL. 

5.1.6 Issue 6: Incidents and Special Activities need to be Better Referenced 

There are no references to reports on the specific incidents or special activities that could be used 

in reconstructing doses for individuals involved in them. The list of special incidents or 

activities on page 12 of the TBD is quite brief. Table 3.1 of the PACE/UTAH (2000) report 

(reproduced here as Table 5.1-1) gives a better chronology of events to be considered in dose 

reconstruction. The Phase II report (DOE 2000) identifies many incidents at the plant from the 

beginning of plant operations up to the early 1990s (PACE/Utah 2000). Are there any 

documents or databases where incident reports are maintained? If so, reference to those reports 

would be helpful. 

5.1.7 Issue 7: The TBD Needs to Consider Operations Other than Gaseous Diffusion 

There is no mention of the smelting operations that took place in Building C-746B. In addition 

to smelting, the diffusion barriers removed from PGDP and other enrichment facilities during 

cascade improvement and upgrade programs, and special smelting projects were also performed 

for other facilities. These activities could potentially affect worker doses. 

PGDP participated in the “Work for Others” (WFO) program providing assistance to other DOE 

sites. There is no mention of this program in the TBD and how this may affect the dose 

reconstructions in terms of radionuclides. Materials and equipment were sent to PGDP from 

other DOE sites, such as the Savannah River Site (SRS), K-25, the Y-12 Plant, Hanford, Los 

Alamos National Laboratory (LANL), Lawrence Livermore National Laboratory (LLNL), 

Sandia National Laboratory (SNL), the Navy, and the National Aeronautics and Space 

Administration. For example, one such project involved the disassembly and subsequent 

recovery of precious metals from weapons components. Much of the WFO information is 

sensitive and would likely require a classified review. 

5.1.8 Issue 8: The Intended Use of the Bechtel-Jacobs Radionuclide Concentrations is not 

Clear 

Pages 11–20 of the TBD provide maximum 237Np, 239Pu, and 99Tc concentrations expected for 

activities at each of the buildings related to the processing of recycled uranium. These relative 

mass concentrations (BJC 2001 does not associate specific radionuclides with Np or Pu) are 

associated with the types of exposures associated with worker activities in that building. The 

reference mass concentration in each case is that of the item shown in parentheses, e.g., MgF2, 

ash, UF6, etc. Bechtel-Jacobs (BJC 2001) uses these values, along with other data, to classify the 

radiation exposure potential for the worker activities. The TBD has misinterpreted the 

classification category as applying only to external radiation exposure potential. SC&A notes 
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that the use of these data for calculation of internal doses would not necessarily be consistent 

with the methods used in the Occupational Internal Dose TBD (Berger 2004). 
 

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Table 5.1-1. Overview of the PGDP Chronology with Emphasis on Events that may have 

had Some Radiological Consequences 

(Source: PACE/Utah 2000, Table 3.1) 

Date Company Event 

Oct 1950 Union Carbide PGDP Site selected 

July 1952 Uranium received 

Sept 1952 Cascade Buildings C-331 and C-333 begin operation 

1953 Aware that feed from recycled reactor fuel contains trace quantities of plutonium 

Sept 1953 2 ash handlers’ urinalyses test positive for Pu 

July 1954 Cascade buildings C-335 and C-337 begin operation 

Aug 1954 First cascade improvement started, plant remains in operation 

Aug 1956 C-420 feed plant online 

Nov 1956 C-310 fire 

Jan 1957 Begin producing U metal and UF4 from depleted uranium 

1957 Radiochemical analysis confirms presence of neptunium and plutonium (separate 

radiochemical analysis indicates entire cascade is contaminated with Np) 

Nov 1958 Neptunium recovery started from receiver ash and cylinder heals 

Apr 1960 Technetium recovery begins 

June 1961 First cascade improvement completed 

Sept 1961 MgF2 traps installed to capture Np in effluents 

Mar 1962 C-340 explosion and fire due to burnout of Mg bomb where uranium was released 

to the furnace 

Mar 1962 Neptunium recovery ends 

Dec 1962 C-337 Explosion and fire 

Jan 1963 Technetium traps installed 

June 1963 Technetium Recovery Ends 

Apr 1968 Radiation overexposure to two maintenance workers* 

Mar 1973 Second cascade improvement started 

Oct 1973 No longer produces U metal; still produces UF4 as a byproduct of HF 

Recovery 

Jan 1975 NRC and ERDA assume AEC 

Mid 1970’s Tc removed as an environmental protection measure 

May 1977 Feed plant ceases operation, receive feed as UF6 

Oct 1977 DOE assumes ERDA 

Jan 1978 C-315 explosion and fire 

Sept 1981 Second cascade improvement completed 

Apr 1984 Martin Marietta 

June 1990 Tiger Team Assessment 

1992 USEC established 

July 1993 Lockheed Martin 

July 1993 USEC (United States 

Enrichment Corporation) 

Leases enrichment production facilities as operations and maintenance contractors 

Nov 1996 NRC grants certificate of compliance to USEC 

Apr 1998 Bechtel-Jacobs DOE clean-up contractor 

May 1999 USEC Assumes direct operation of enrichment facilities 
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5.2 OCCUPATIONAL MEDICAL DOSE 

Background & Introduction 

The PGDP TBD for Occupational Medical Dose, ORAUT-TKBS-0019-3 (Turner 2004, 

Section 3), makes reference to occupational dose guidelines in Kathren 2003 as its basis 

document for estimating medical dose. Kathren 2003 has since been revised as Dose 

Reconstruction from Occupationally Related Diagnostic X-ray Procedures, Revision 3 (Kathren 

and Shockley 2005), which provides a more reasonable basis for assumptions regarding 

estimation of worker medical exposures at Mound. The SC&A review evidences that the TBD 

recognizes the total lack of exposure data and protocols existed prior to 1975. In Table 3-3, the 

TBD details technique factors for x-rays prior to 1975, which are based solely on assumptions 

with no documentation provided. NIOSH did conduct interviews with PGDP Medical staff to 

ascertain any personal knowledge on x-ray procedures. 

Specific dose estimations in the TBD for occupational x-rays are also derived for the period prior 

to 1975 from ICRP Report 34 (ICRP 1982) and NCRP Report 102 (NCRP 1989). This is an 

important issue in that estimated medical doses based upon this research are not actual 

measurements of dose from the PGDP x-ray unit. For the period of 1995–present, it is stated that 

there were measurements made by the State of Kentucky and the Food and Drug Administration 

(FDA); however, no references are provided. 

The TBD states there is no physical evidence to show if photofluorography (PFG) to do chest 

screenings occurred at PGDP. In the absence of documentation, the TBD directs that dose 

assessors not include the use of PFGs in their dose estimations. This direction contradicts OTIB0006, which recommends assuming annual fluoroscopic examinations in the early years lacking 

evidence to the contrary. 

Additionally, the TBD does state that all medical x-rays administered in conjunction with routine 

or special exams are considered as part of the occupational exposure; however, only preemployment and routine chest exams are mentioned to be used by dose assessors. Review of the 

subject TBD has resulted in a number of issues and supportive observations, which are important 

to NIOSH to consider and clarify, due to the risk inherent to medical exposures, to assure that 

considerations are claimant favorable. These issues and observations are discussed below. 

5.2.1 Issue 1: The TBD fails to Adequately Define and Assess Occupational Medical 

Exposure 

The guidelines as presented in ORAUT-OTIB-0006, Revision 2 (Kathren 2003), go a long way 

in assuring that all occupational medical exposures are reasonably included in determining the 

overall dose estimations for claimants. Unfortunately, the interpretation to date by the contractor 

(ORAU) has not been applied conservatively to be claimant favorable. The Occupational 

Medical Dose TBD (Turner 2004) assumes an interpretation that has been also considered and 

applied at other sites, such as the Mound Plant and Los Alamos National Laboratory (LANL). 

To this extent, the assumption that medical procedures are limited to only one pre-employment 
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chest x-ray and chest x-rays that are part of routine physical exams, may substantially 

underestimate worker medical exposure, when evaluating occupational medical exposure. 

In more recent documentation, Revision 3 of ORAUT-OTIB-0006 (Kathren and Shockley 2005) 

concluded that other examinations should be included, such as special screening exams (e.g., 

respiratory protection, beryllium workers, asbestos workers, etc.) and termination exams. The 

occupational medical TBD does not recognize this change from the previous Revision 2 of the 

TIB, and also assumes that special chest radiography for respirator certification, beryllium and 

asbestos workers, and food handlers are accomplished as part of the routine physicals. This is 

not documented in the medical TBD. Another factor not discussed in the TBD is the potential 

impact of x-ray procedures utilized by medical authorities to do special screenings that are 

performed outside the frequency listed in Table 3-1 of the TBD. 

The TBD (Turner 2004) concludes that chest examinations are often quite limited after 1974. 

Only smokers are documented in Table 3-1 of the TBD as having received an exam every 

3 years. The only differential noted is that asbestos workers after 1986 had a chest exam 

performed every 2 years. To the contrary, there is ample evidence that annual chest x-rays were 

often provided on a voluntary basis to nearly all workers. The majority of workers had chest 

x-rays each year as a routine at DOE sites until the mid-1980s, when Federal guidelines warning 

against routine screening were first being enforced. 

After discussion with NIOSH personnel, it was their decision to limit occupational medical 

exposure to those chest exams described above, and to conclude that all other exposures are part 

of worker background. SC&A believes such an interpretation is not claimant favorable to those 

most at risk. Our concern is that specified “high-risk” workers, those most likely exposed to 

radiation and beryllium, would be at risk of having an incomplete dose assessment if not all 

radiation associated with medical screening for job-related activities were included. Since all 

radiation provides some risk and, arguably, is cumulative, workers warrant consideration of all 

forms of work-related x-ray exposure to be claimant favorable. SC&A believes NIOSH should 

review its interpretation of included medical exposure, and should reasonably adopt a broader 

interpretation of occupational medical dose, as provided in the most recent version of ORAUTOTIB-0006 (Kathren and Shockley 2005). This is particularly important for skin cancer, which 

is not one of the SEC-defined cancers. 

5.2.2 Issue 2: Techniques and Protocols Increase Uncertainty of DCFs listed in the TBD 

Sections 3.2 and 3.3 of the TBD (Turner 2004) fail to describe information upon which to 

establish beam quality for x-ray units in use from 1952–1975. In 1975, the site installed a singlephase Picker unit. There is only limited documentation to show that the Picker unit, in use from 

1975 through 1995, had added filtration, as first measured by the FDA and the State of Kentucky 

after 1995. In the absence of definitive tube output measurements, the TBD directs the use of 

default values and dose conversion factors (DCFs) derived from ICRP Report No. 34 (ICRP 

1982). These values are then applied to determine organ doses using Tables A.2 through A.8 of 

ICRP Report No. 34 (ICRP 1982). An issue of concern is that the DCFs are derived using a 

default half value layer (HVL) of 2.5 mm aluminum (Al) for Type 1 units, in use from 1946– 
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1980. Another issue is that the unit, in use from 1975–1995, had an unknown added filtration of 

Al, which may be different from the default value from NCRP Report 102 (NCRP 1989). 

The Occupational Medical Dose TBD (Turner 2004) provides little documentation to support the 

assumed techniques and protocols applied to calculate the dose, which is mainly derived from 

NCRP Report 102. The TBD states that a posterior-anterior (PA) chest x-ray was typically the 

only view. An undocumented assumption in the TBD is that exams required only a PA view. 

SC&A has inquired whether a definitive protocol existed to validate that chest exams included 

PA views and lateral (LAT) views only on a limited basis after 1975. NIOSH has acknowledged 

that the lack of verifiable protocols is a generic problem at many sites, has planned to search all 

available records, and will include pertinent records and references in any future revision of this 

section of the TBD. 

The Occupational Medical Dose TBD is also deficient in that little documentation exists to 

validate x-ray protocols and equipment maintenance, and upkeep of records prior to 1995, after 

which the State of Kentucky commenced surveys of the unit. The TBD uses information derived 

from the TIB (ORAU 2003) to estimate dose impacts. 

5.2.3 Issue 3: Frequency and Type of X-ray Exposure Is Uncertain 

The Occupational Medical Dose TBD relies on a very limited review of current medical 

personnel to establish frequency assumptions. The assumption of one chest radiograph (PA) 

every 3 to 5 years is not reasonably conservative, in that workers could request an x-ray, or be 

subject to special screening exams. The frequency of screenings, and the number and type of 

workers receiving x-rays varies from site to site. 

The Occupational Medical Dose TBD in Section 3.2 provides no documentation or references to 

support the assumption that only a limited group of workers received x-ray exams more 

frequently than every 5 years after 1952. To the contrary, up until about 1985, most DOE sites 

performed chest x-rays almost on a voluntary basis. DOE medical program reviews documented 

during the early 1990s showed many sites still used chest radiography as a general screening 

exam. Most workers accepted chest x-rays, even though the job did not require it. Also, the 

assumption that workers in special exposure categories, such as beryllium workers, were given 

chest x-rays only as part of their routine physical is not well-documented and not consistent with 

special screening guidelines. The TBD applies no conservative assumption to cover such exams. 

Section 3.2 of the Occupational Medical Dose TBD states that PFG units, although generally 

available at most DOE sites up to the late 1950s, were not used at the PGDP. The undocumented 

absence of PFG units at PGDP clearly has significant dose implications to workers who may 

have been given much higher doses from PFG units. The PFG unit provides a dose to the worker 

greater by a factor of 5–6 more than that delivered by conventional radiography. The TBD does 

not provide documentation for the types of equipment in use at PGDP prior to 1995. SC&A 

believes it is not claimant favorable to instruct dose assessors to use kerma (dose) values of 200 

mrem and 100 mrem for chest radiography prior to 1975. To be fully claimant favorable, it 

would be appropriate to instruct dose assessors to use an annual dose of 3.0 rem per year for 
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chest radiographs, in accordance with guidelines set forth (Kathren and Shockley 2005) until 

1959, when you can assume all PFGs were no longer used at DOE sites. 

5.2.4 Issue 4: Other Potential Medical Exposures have not been Identified 

The Occupational Medical Dose TBD (Turner 2004) does not address the potential use of forms 

of radiation exposure other than x-ray units to support medical diagnosis. This may involve use 

of isotopes, sealed sources, etc. The TBD is also deficient in that it does little to catalog the 

number, types of x-ray equipment, frequency of use, etc., discussed above in Issues 2 and 3. 

The below-average performance at the PGDP to perform routine and preventative maintenance 

during the 1952–1975 timeframe suggests that routine maintenance of x-ray units is not likely, 

unless performed by an unknown outside contractor. Unfortunately, no records exist to evidence 

maintenance, calibrations, etc. The lack of defined protocols and basis for approval of 

radiography procedures suggests that the use of radiography was not closely controlled. The 

Occupational Medical Dose TBD does not discuss the use of portable radiography to perform 

screenings and the potential for exposure of medical personnel or other workers without 

dosimetry devices being utilized. Interviews with medical staff suggest that portable x-ray units 

were not used. This is potentially an issue for the PFG unit, which was often van-mounted at 

other sites. Additionally, the TBD fails to document that available x-ray units were not operated 

at greater than 80–90 kVp prior to 1975. To the contrary, Table 3-3 of the TBD indicates that 

the kVp after 1975 was set at 100 kVp when performing LAT chest exams. 

The conclusion is that the TBD does little to reasonably document the variety of medical 

occupational exposures, and the lack of documentation on the type of equipment and the 

maintenance records do little to assure that a conservative and claimant-favorable estimation of 

dose is possible. This circumstance would suggest the need to reconsider a worst-case approach 

to establishing dose. NIOSH should revisit and update Sections 3.2 and 3.3 of the TBD as 

needed. 

5.2.5 Issue 5: There are Additional Factors that Contribute to Uncertainties 

The Occupational Medical Dose TBD does not consider dose impacts due to less-than-optimal 

use of technology, such as using screens, grids, or bucky systems. The TBD does not consider 

these elements as potential contributions to uncertainty. 

The TBD does consider the potential contribution to dose that may have resulted in less than 

optimal use of collimation, at least prior to 1975, as stated in Section 3.5 of the TBD, and offers 

substitute DCFs for use by dose reconstructors for selected exams, as presented in Tables 3-6 and 

3-7. Unresolved is the concern that the DCFs are derived from ICRP (1982), and therefore, are 

not comparable in terms of beam quality, which varies from unit to unit. These factors can 

contribute greatly to the dose to the chest and other organs for units in operation prior to 1995, 

where little or no documentation exists. NIOSH has indicated that it will continue to search for 

other available records to better define equipment use and beam quality, and include it in an 

updated version of the TBD, as appropriate. 
 

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Uncertainty is defined in the TBD as being due to measurement error, and variation in 

kilovoltage, tube current, timers, and the source-to-surface distance (SSD). This approach is 

quite similar to the uncertainty analyses documented in other DOE site profiles. The conclusion 

in this TBD, and others, is that an uncertainty factor of +30% should be used by dose 

reconstructors for exposure prior to 1975. SC&A believes the correction factor of 2.0, being 

applied after 1975, is more appropriate. 

SC&A agrees that the TBD conservatively estimates these essential aspects of an uncertainty 

review. Unresolved is the contribution to uncertainty in dose, due to other errors introduced by 

lack of quality controls in processing equipment and lack of adherence to established SOPs. A 

reasonable estimate of these contributions to uncertainty would be an evaluation of retake rates 

per examination type. NIOSH should revisit the potential for significant retake rates and 

evaluate its potential effect on dose as part of future revisions of this TBD, especially as it relates 

to examinations prior to 1975. 

The Occupational Medical Dose TBD does not show that PGDP applied dose minimization 

principles to reduce medical exposures. The document also does not assess or consider the likely 

exposure to workers who are referred to off-site medical facilities for follow-up. The TBD states 

that review of selected medical records and files did not reasonably show or match expected 

x-ray exam frequency, and type of exam. Little evidence exists to document the number of x-ray 

exams provided to the average worker, or for special exposure needs. 

5.3 OCCUPATIONAL ENVIRONMENTAL DOSE 

Background & Introduction 

The TBD for the PGDP Occupational Site Description, ORAUT-TKBS-0019-4 (East 2004), 

describes the potential exposures from ambient sources to workers while outside the process 

buildings. The document presents data for estimating annual intakes from radionuclides in air 

and external dose from ambient radiation, as well as from cylinder storage yards. 

The document does not present radionuclide-specific measurements. Uranium concentrations 

are inferred from measurements of alpha concentrations in air and Tc-99 concentrations from 

beta concentration in air measurements. Until September 1981, external gamma radiation levels 

were monitored with Geiger Mueller counters. Since then, thermoluminescent dosimeters 

(TLDs) have been used for that purpose. The TBD provides references to numerous 

environmental reports for the site. 

5.3.1 Issue 1: Onsite Environmental Exposures are based on Site Boundary Data 

The TBD purports to describe “potential exposures from ambient sources while working outside 

the process buildings,” but the ambient monitoring data in the document are for measurements at 

the site boundaries and beyond. There are no corroborating data provided to demonstrate that 

these measurements are consistent with the levels that workers might experience while working 

at the site outside the process buildings. 
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5.3.2 Issue 2: There are No Comparisons between Predicted and Measured 

Environmental Exposures 

The TBD states that since 1962, “At PGDP all personnel wore film badges....” This document is 

supposed to provide information for estimating environmental doses when monitoring was not 

performed or coworker data could not be used, but there are no comparisons between doses 

based on personnel badge data with estimates based on ambient environmental exposures. Such 

comparisons would help to validate the methods for reconstructing environmental doses. 

5.3.3 Issue 3: The Pre-operational Measurements could not have detected Ambient 

Levels of Uranium in Air 

East (2004) states that none of the 10 pre-operational samples analyzed in 1952 showed any 

measurable concentration of uranium in air (0.00 mg/m3

), and only one of the two alpha activity 

samples showed any measurable alpha activity in air (0.018 cpm/ft3). There are no comparisons 

with other ambient measurements. ATSDR (1999) published a table of uranium activity in air 

values measured in 1984 and 1993 at over 30 sites. The total uranium activities (234U + 235U + 

238U) in these measurements range from about 4.6 aCi/m3

 (Syracuse, New York, 1984) to 

341.4 aCi/m3 (Lynchburg, Virginia, 1984). An alpha concentration of 0.018 cpm/ft3 (assuming a 

100% counting efficiency) would correspond to about 290,000 aCi/m3 if all the alpha radiation 

came from natU—about 850 times the Lynchburg measurement. Clearly, the one positive alpha 

activity cannot be interpreted as due to the pre-operational concentrations of uranium in air. On 

the other hand, if the MDL for uranium mass concentration measurement was about 0.01 mg/m3 

(10 μg/m3

), that would correspond to a natU activity concentration of about 6.9 pCi/m3, a level 

about 20,000 times higher than the Lynchburg measurement. In conclusion, it would appear that 

neither of the methodologies used for determining pre-operational levels of uranium in air at the 

site was adequate for that purpose. 

5.3.4 Issue 4: The Ambient Air Sampling Collection Network was not Intended for 

Monitoring Onsite Concentrations 

The four on-site air monitoring stations located at the security fence (according to Table 4-1 of 

the TBD) were primarily intended for demonstrating compliance with offsite air quality criteria. 

There is no map showing the locations of the air sampling stations in the network. 

5.3.5 Issue 5: Environmental Dose Gaps Exist in the TBD 

It is not clear whether or not air transport modeling was performed for the site. There are a 

number of considerations identified that limit the use of air transport modeling to estimate 

airborne concentrations from site releases, but there are no comparisons made between the site 

boundary data and any such modeling that had been done. 

There is little characterization of the releases in Tables 4-2 and 4-3. There is no discussion of 

what release points were considered or the physical form of the uranium and 99Tc release data in 

Tables 4-2 and 4-3, other than to indicate that they were obtained from DOE, Bechtel-Jacobs, 

and United States Enrichment Corporation reports. 
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Section 4.3, External Dose, does not consider the radionuclides in the depleted uranium cylinders 

providing the gamma radiation for the increasing external exposure rates near the depleted 

uranium cylinder storage yards. SC&A questions the use of TLDs as environmental dosimeters 

starting in 1961. That year seems rather early for the use of these TLDs. Early survey results of 

about 0.02 mrem/hr (probably measured as 20 μR/hr) would correspond to 40 mrem in a 

2,000-hr period or about 175 mrem/yr for continuous exposure. For comparison, the average of 

the 1993–2001 reported background measurements in Table 4-4 correspond to a 100 mrem/yr 

rate for continuous exposure. Presumably, the earlier survey was in an era when fallout from 

weapons testing would have contributed significantly to ambient levels. 

Section 4.4, Uncertainty, deals less with estimating the uncertainties in measured or calculated 

values than with outlining default values and assumptions to be used when data are unavailable. 

5.3.6 Issue 6: Comments on Section 4.3.2, Radiological Conditions in Cylinder Storage 

Yards 

(1) The statement that dose rate measurements at 100 m from the UF6 storage yards are 

representative of dose rates near the storage yards is not supported with measurements. It 

would seem likely that dose rates could increase substantially as the cylinders are 

approached. 

(2) The statement, “During recent years this area has been posted as a radiological area, 

which has reduced the number of unmonitored workers spending any significant time in 

the area to zero,” raises some questions: 

a. What were the criteria for the designation? 

b. Given the 1962 policy that all workers were to wear dosimetry at all times, who were 

the unmonitored workers this policy was designed to protect? Were they workers 

whose badges were not normally analyzed? 

5.3.7 Issue 7: There is Little Characterization of the Effect of Plant Releases on the 

Onsite Levels of Exposure 

There is no discussion in the TBD of radionuclide-specific concentrations in air, soil, or water. 

There is no discussion of the environmental fate of uranium and other radionuclide releases. In 

effect, there is an implicit a priori assumption that the dosimetry systems in use at PGDP will 

adequately reflect any environmental exposures. Two additional potential sources of ambient 

exposure include the burning of contaminated material in onsite pits and the alleged intentional 

releases that occurred in the 1950s. Further investigation into these areas is necessary. 

5.4 OCCUPATIONAL INTERNAL DOSE 

We have reviewed the Technical Basis Document for Paducah Gaseous Diffusion Plant – 

Occupational Internal Dose (Berger 2004), as well as Internal Dosimetry Coworker Data for 

Paducah Gaseous Diffusion Plant (Ikenberry 2005), and have identified a number of key issues, 
 

 

 

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which are presented below. The next part of this section presents a detailed discussion of these 

issues, keyed to the sections and tables in the TBD.2 

5.4.1 Key Issues 

Issue 1: Completeness of Data Sources—Inadequate Characterization of the Source Term 

A critical issue is the specification of the source term for the internal exposures. This TBD does 

not fully utilize information presented in two key documents. One is the exposure assessment of 

workers at the PGDP (PACE/Utah 2000), which emphasized exposures to neptunium and 

plutonium. Although this report is cited as a source of information on a few narrow topics, the 

extensive data summarizing the results of radiological surveys were not incorporated in 

developing the default specific activities, which are needed for reconstructing the doses to 

workers whose bioassay results were limited to total concentrations of uranium. 

The other document is the draft report on recycled uranium mass balance (BJC 2001), which is 

cited as a source of information by Turpin (2006), but is not mentioned in this TBD. That report 

breaks down the potential exposures to three contaminants in recycled uranium—99Tc, 237Np, 

and plutonium—into 11 types of activities and operations at the PGDP, and lists the maximum 

concentrations of each of these constituents (BJC 2001, Table 2.4-1). Other data, such as those 

in DOE 2000, are erroneously characterized. 

Issue 2: Inconsistencies and Inaccuracies in the Source Term 

The TBD is inconsistent in assigning radionuclides to the various facilities at the PGDP. Notable 

among these are the omission of transuranics (TRU) in the feed plant and the “cascades 

(general)” facilities. 

The default specific activities of nuclides other than 235U and 238U have been underestimated by 

as much as 5 orders of magnitude. Prior to 1989, most urinalyses in the electronic database only 

list the chemical assays of uranium. Information on the specific activities of the uranium 

isotopes is therefore required to estimate intakes of the activities of each isotope, which are 

needed for dose reconstruction. The specific activities of other radionuclides (i.e., activity of 

each nuclide per unit mass of uranium) are required to estimate the intakes of these nuclides, 

based on the mass concentration or daily excretion of uranium in urine. 

A more detailed discussion of issues involving the source term is presented in Section 5.4.2 of 

this review entitled “Section 5.2: 'Source Term'.” 

Issue 3: Particle Size and Lung Clearance Types 

The TBD assumes all respirable aerosols have an AMAD of 5 μm, despite evidence of smaller 

particle sizes. Such an assignment is, in most cases, not claimant favorable. Incorrect Lung 

2

 In the remainder of this section, the term “TBD” is used to refer to the “Technical Basis Document for Paducah 

Gaseous Diffusion Plant – Occupational Internal Dose,” ORAUT-TKBS-0019-5. 
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Clearance Types are assigned to the various nuclides in many facilities. The assigned types are, 

in many cases, not claimant favorable. 

Issue 4: Inadequate and/or Inaccurate Guidance on Estimating Intakes from Bioassay 

Results 

The TBD presents inadequate, and sometimes inaccurate, guidance to dose reconstructors on 

estimating the intakes of uranium, including the assumed specific activities of other 

radionuclides not analyzed by the bioassays. The issues include making claimant-favorable 

assumptions about the time of collection of urine samples with respect to the work week, and the 

utilization of “recall” samples in cases the original assay indicated a high intake of radioactive 

materials. A more detailed discussion of issues involving the estimation of intakes is found in 

Sections 5.4.3 and 5.4.4 of this review entitled “Section 5.3: 'In-Vitro' Measurement Methods” 

and “Section 5.4: 'In-Vivo' Measurement Methods,” respectively. 

Issue 5: Erroneous Equations for Estimating 24-hour Excretion 

The equations for estimating 24-hour excretion on the basis of spot urine samples are incorrectly 

written. There is an error in the presentation that can lead to errors in the interpretation and 

application of these methods. Furthermore, the daily excretion of urine is based on obsolete data. 

Issue 6: Invalid Guidance for Estimating Intakes Based on Time of Sample Collection 

The guidance for utilizing the time of sample collection in dose reconstruction, or assuming a 

time if none is listed in the records, is neither scientifically valid nor claimant favorable. Large 

variability in the calculated intakes can result if the time of collection following exposure is not 

properly accounted for. 

Issue 7: Omission of Guidance on Use of Breath and Fecal Analyses 

Limited fecal analyses, and possibly breath analyses, were performed at PGDP. The TBD 

provides no information on such bioassays. 

Issue 8: Inadequate and Internally Inconsistent Discussion of In-Vivo Measurements 

Although in-vivo bioassays were not widely performed at the PGDP, the TBD should 

nevertheless present adequate information to allow dose reconstructors to properly utilize any 

data that may be available. The discussion of in-vivo measurements and MDAs is inadequate 

and internally inconsistent. 

Issue 9: Inadequate Discussion of Significant Incidents and Accidents 

The information on significant incidents and accidents is extremely sketchy and of little use to 

dose reconstructors. 
 

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Issue 10: Unsuitable Use of Coworker Models 

According to a more general guidance document on the use of coworker bioassay data (Brackett 

2005), the basis for the application of a general coworker model is as follows: 

... participation in a bioassay program involves workers who have the largest 

potential for exposure. While there are exceptions to this generality such as 

accidents involving unmonitored workers, it is unlikely that an unmonitored 

worker would have received a larger dose than the most highly exposed 

monitored worker at a site. 

However, Brackett provides no statistical or quantitative technical basis for this hypothesis. 

In the coworker models used by NIOSH, workers are not classified by their jobs or by the 

buildings where they performed their work. The model described by Ikenberry does not identify 

jobs that may have led to a higher probability of intake by workers, either in routine or special 

circumstances. Air sampling data from the PGDP show differences in concentrations of uranium 

dust in the air in the various buildings, which would lead us to expect that workers in buildings 

that had higher concentrations of uranium in the air would have higher concentrations in the 

urine. However, there was no attempt in this report to sort the urinalysis results by job 

assignment or location to determine if there were any correlations between the uranium 

concentration in the urine and the building or department where the job was performed. 

In order to apply the coworker model to unmonitored workers, NIOSH needs to demonstrate that 

there is a low probability that any unmonitored worker could have higher exposures than the 

monitored workers taken as a group. Ikenberry fails to do so. 

In fact, there is evidence that the monitoring program was designed to encompass all workers. 

According to PGDP (2003a): 

The frequency of such industrial urine samples for uranium varies from a 

maximum frequency of each four weeks for all personnel working routinely in 

C-310, C-315, C-340, C-400 and C-4l0 to a minimum frequency of each 

12 months for those working in some locations where there is little likelihood of 

any exposure to this element or its compounds. 

The above statement negates the assumption that the unmonitored workers had lower 

probabilities of radiation exposures. If all workers were supposed to be monitored, unmonitored 

workers might well be those whose records were lost or who, for some unknown reason, did not 

participate in the program. 
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Issue 11: Use of Unverified Bioassay Data 

Dodd (2002) states the following: 

The data that is included in the 

“PGDP_Historical_Internal_Data_(Unverified).mdb” database consists of invivo data and urine card data. This [sic] data has not been through a verification 

and validation process; therefore it should not be considered data of record. 

This caveat raises a question about the use of the cited database, which is the basis of bioassay 

data for the 1952–1976 period. This issue needs to be addressed by NIOSH before such data can 

be used for the coworker model—in fact, before they can be used for any dose reconstruction. 

Issue 12: Incorrect Calculation of Intakes from Mass Concentrations of Uranium in Urine 

According to Ikenberry: 

The total uranium results are in units of μg/L; therefore the results were 

multiplied by 1.4 in order to normalize them to the Reference Man excretion rate 

of 1400 mL per day. Bioassay results were converted from mass to activity before 

fitting assuming 0.0389 Bq/μg, characteristic of low-enrichment (2 percent) 

uranium. 

The current ICRP (2002) reference values are 1.6 L/d for an adult man and 1.2 L/d for an adult 

woman. Therefore, the TBD should be revised to reflect these values. Furthermore, the default 

specific activity of uranium, based on the measured concentrations of uranium isotopes in the 

PGDP cascade product, should be 0.0541 Bq/μg, instead of 0.0389 Bq/μg, as cited by Ikenberry. 

Issue 13: Lack of Documentation for Choice of Distribution Type 

Brackett (2005) describes the statistical analyses for developing a site-specific coworker 

bioassay model. These include calculating the goodness-of-fit to determine whether the bioassay 

results conform to a lognormal distribution. Finally, according to Brackett: 

All decisions regarding the statistical analysis are recorded in instructions to the 

statistics analyst and in a site-specific document such as the site profile or a TIB. 

Ikenberry states that the urinalysis results for individual calendar-year quarters have lognormal 

distributions. However, he cites no statistical results that support this conclusion; furthermore, 

the TIB fails to indicate whether any such statistical analyses were performed. Absent such data, 

the compilation of 50th and 84th percentiles have little meaning. 
 

 

 

 

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Issue 14: Incorrect Selection of Distinct Time Periods 

According to Ikenberry: 

The analysis started a new chronic intake period if the data indicated a 

significant sustained change in the bioassay results. By this method, the period 

from 1952 through 1988 was divided into two chronic intake periods. 

The author gives no detailed explanation of how the division into two time periods was 

accomplished. Our own examination of the data, followed by a statistical analysis, finds that 

there are three distinct time periods, as shown in Table 5.4-1 of this review. 

Table 5.4-1. Urinalysis Results (Bq/d) for Three Distinct Time Periods 

Time Period 6/1/1952 – 3/31/1959 4/1/1959 – 3/31/1980 4/1/1980 – 12/31/1988 

Number of quarterly results 24 84 35 

Mean 0.398 0.282 0.084 

50th percentile S.D. 0.084 0.078 0.023 

S.E. 0.017 0.009 0.004 

Mean 1.212 0.744 0.256 

84th percentile S.D. 0.521 0.197 0.064 

S.E. 0.106 0.022 0.011 

As shown on the table, the three time periods exhibit mean values of both the 50th and 84th 

percentiles that are significantly different from each other, as shown by the standard error of each 

mean, which is far smaller than the difference between the means. Thus, the three time periods 

encompass three distinct sets of values. 

Issue 15: Coworker Data Inapplicable to Individual Unmonitored Workers 

The applicability of the coworker data presented by Ikenberry to individual unmonitored workers 

is questionable on several grounds. The models presented by the author represent a best estimate 

of the constant, chronic intake for each of two time periods by a hypothetical individual whose 

quarterly urinalyses results correspond to the median values of all workers for a given quarter. A 

similar estimate is made for a hypothetical worker whose urinalysis results correspond to the 

84th percentile values for each quarter. There are three sets of values—one for each uranium 

Lung Clearance Type; F, M, and S. A distribution of uranium intakes for each time period and 

each Lung Clearance Type was constructed. The intake of the hypothetical median individual 

was assigned to the median value of this distribution, while the ratio of the intake of the 

hypothetical 84th percentile individual to that of the median individual was assigned to the 

geometric standard deviation (GSD). If the latter value is less than 3, the GSD is assigned a 

value of 3. 
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While this distribution might well represent the intakes of the population of monitored workers, 

there are several reasons why it cannot be used as a surrogate for the intake of an unmonitored 

worker: 

 The data for this distribution includes assays of workers with very low potentials for 

exposure. According to PGDP (2003a, see excerpt quoted previously), all workers were 

to submit urine samples at least once every 12 months. Consequently, the distribution 

would not be claimant favorable for workers with job assignments that would have 

subjected them to significant radiation exposures. 

 The model assumes a constant rate of intake; however, the plots of the 50th percentile 

excretion rate of uranium show considerable variability among the quarters within a 

given time period. The 84th percentile rates are even more variable. Thus, the 

assumption of a steady intake is not supported by the data. 

 The intake rate is based on a steady exposure during the entire time period. It is unlikely 

that a given unmonitored worker would have been continually exposed during, say, the 

period from June 1, 1952, to March 31, 1980—a period of 28 years. 

 As pointed out earlier in this review, urinalyses during the period from June 1, 1952, to 

March 31, 1959, had significantly higher values than those between April 1, 1959, and 

March 31, 1980, yet they were combined into a single period. A worker exposed during 

the earlier period would thus be assigned a lower intake rate than had the rates for the 

three periods been calculated separately. 

In conclusion, the lognormal distributions described by Ikenberry are artificial constructs that are 

not based on any real individuals. The application of such distributions to unmonitored workers 

is neither scientifically valid nor claimant favorable. 

The remainder of Section 5.4 contains a detailed discussion of the TBD, keyed to the sections 

and tables in that document. 

5.4.2 Section 5.2: “Source Term” 

The first paragraph of Section 5.2: “Source Term,” singles out 230Th and 234mPa as uranium 

progeny “of dosimetric interest.” This designation of 230Th is apparently an error; DOE (2000) 

cites 234Th and 234mPa, the short-lived progeny of 238U. According to DOE, these nuclides are 

significant, because when the uranium is further processed, significant quantities of these 

daughter products can remain behind in the form of oxides or ash or on the surface of process 

vessels. Most of the beta dose from uranium is, in fact, due to these two radionuclides. 

Consequently, high concentrations of the nuclides would produce much higher fluxes of beta 

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5.4.2.1 Table 5-1: “Elemental Fractions of Uranium, Neptunium, and Plutonium at 

Specific Facilities” 

Table 5-1 of the TBD lists activity fractions of uranium, neptunium, and plutonium at several 

facilities of the PGDP; the report recommends that these values be used by dose reconstructors. 

First, as the author acknowledges in footnote “a” to this table, the title, “Elemental fractions of 

uranium, neptunium, and plutonium at specific facilities,” is misleading. Despite how the data 

appears in the source document, it would be better to use the term “activity fractions” rather than 

“elemental fractions,” if this is indeed what the values represent. 

More important, the fractions of neptunium and plutonium at the feed plant (Buildings C-410 

and C-420) and at “Cascade (general)”3 are listed as zero. This ignores the data in Table 7.9 of 

PACE/Utah 2000, which lists activity fractions of 0.7% 237Np and 9% 239Pu in the C-410 control 

room and fluorination tower air concentrations, and lower fractions in other feed plant locations. 

It is also inconsistent with Table 5-5 of the TBD, which lists 237Np and 239Pu in Buildings C-410 

and C-420. These activity concentrations can have large dosimetric consequences, given that the 

50-y dose to a non-metabolic organ (e.g., testes) from unit intakes of Type M 237Np or 239Pu are 

more than two orders of magnitude greater than the comparable dose from 238U. 

Table 5-1 does indicate neptunium and plutonium activities in Buildings C-333 and C-337, 

which housed portions of the cascades. However, there is no mention of Buildings. C-331 and 

C-335, which also contain parts of the cascades. This is again inconsistent with Table 5-5, which 

lists 237Np and 239Pu in Buildings C-331 and C-335. 

5.4.2.2 Table 5-2: “Isotopic Fractions for Various Enrichment Percentages” 

The TBD advises dose reconstructors to use Table 5-2, which lists “isotopic fractions” for 

various uranium enrichment percentages, to determine the amount of each uranium isotope when 

bioassay results are listed in terms of the mass of uranium. First, the table is mislabeled; the 

values are specific activities, not fractions. Next, an examination of these data reveals a number 

of discrepancies. The most glaring one is the specific activity of 235U in 93% enriched feed. A 

direct calculation yields an activity of 7.44E-2 Bq/μg, a factor of 10 higher than the listed value 

of 7.38E-3 Bq/μg. 

Smith (1984, Appendix 13) tabulated the measured concentrations of 234U, 235U, and 236U in the 

PGDP cascade product and tails during the period October 2, 1955, to December 31, 1982. 

Using these data, we calculated the mean and maximum concentrations of these isotopes in 80 

cylinders of cascade product withdrawn during this period, along with the corresponding specific 

activities. These results are listed in Table 5.4-2, together with the default specific activities 

from Table 5-2 of the TBD. 

3

 Some further explanation is required of the term “Cascade (general).” 
 

 

 

 

 

 

Table 5.4-2. Concentrations of Uranium Isotopes in the PGDP Cascade Product 

Isotope Mean Concentration Maximum Concentration TBD Default 

Wt % Bq/μg Wt % Bq/μg Bq/μg 

U-234 0.0100 2.26E-02 0.0172 3.89E-02 2.52E-02 

U-235 1.27 9.99E-04 1.95 1.54E-03 1.58E-03 

U-236 0.0072 1.71E-04 0.0625 1.48E-03 3.33E-05 

U-238a 98.72 1.23E-02 97.97 1.22E-02 1.21E-02 

Total 100 3.61E-02 100 5.41E-02 3.89E-02 

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Source: Smith (1984, Appendix 13) 

Note: specific activities are in terms of total uranium

 a Weight percent of 238U calculated by subtracting concentrations of other isotopes from 100% 

When compared to the measured concentrations at the PGDP, the default specific activities of

234U and 236U, listed in Table 5-2, are not claimant favorable. The default specific activity of

234U, while higher than the mean activity, is only 65% of the maximum measured activity, while 

the default 236U activity is much lower than the mean of the measured values. Absent knowledge 

of the specific activities or detailed information on the form of uranium dust to which a worker 

was exposed, NIOSH should assume the maximum measured concentrations listed by Smith, as 

shown in Table 5.4-2, above. 

5.4.2.3 Table 5-3: “Transuranic Element Concentrations in Plant Aerosols (1987)” 

Table 5-3 of the TBD, which lists concentrations of 230Th, 234U, 238U, 237Np, and 239Pu in dust 

samples, is erroneously captioned “Transuranic element concentrations in plant aerosols,” 

although only two of the five radionuclides listed are TRU. The specific activities are listed in 

units of dpm/g U and “nCi/μg U.” The latter heading is incorrect; the units are actually nCi/g U. 

Although this is obviously an editing error, it results in the values being overstated by a factor of 

one million. The values in this table are apparently taken from Baker (1987). In fact, the 

attribution “(1987)” in the title is another editing error—it should read “(Baker 1987).” 

Baker (1987, pg. 6) lists radionuclide concentrations in the UO3 powder area and the 

Building C-420 Green Salt Plant in several different units. There is an inconsistency in the 230Th 

concentration values. The concentration is listed as 1 ppb, which is equivalent to about 

44 dpm/g U, which in turn is equal to a fraction of 2.9 × 10-5 of the total α activity in the dust. 

Baker lists a value of 2.8 × 10-5, which is in good agreement with the value cited above. 

However, he lists a specific activity of 42,000 dpm/g U, which is inconsistent with the other two 

values. This is confirmed by subtracting the activities of 234U, 238U, 237Np, and 239Pu, which are 

reproduced in Table 5-3 of the TBD, from the total activity of 1,500,818 dpm/g U listed by 

Baker. The difference, which is the 230Th activity, is 48 dpm. Thus, the value of 42,000 dpm is 

erroneous, an error that is propagated in Table 5-3. Although this may result in higher doses, it 

is not correct. 

The concentrations in the other areas listed by Baker appear to be internally consistent and are 

correctly reproduced in Table 5-3. However, the table omits information on activities in airborne 

dust at four additional locations or work areas listed by PACE/Utah (2000, Table 7.9); ash 
 

 

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receivers, pulverizer, Building C-400 converter salvage line, and converter maintenance. Two of 

these facilities—ash receivers and pulverizer—have the highest fractions of 230Th and 239Pu in 

the dust. 

Finally, the TBD provides no guidance as to how the data on air concentrations can be used in 

dose reconstruction. Although we agree with the primary reliance on urinalysis, supplemented 

by in-vivo measurements, where such records are available, measured air concentrations could 

be used to supplement the bioassay results when the bioassay records are missing or 

fragmentary. They could be especially useful in assigning relative specific activities of isotopes 

other than uranium. 

5.4.2.4 Table 5-4: “Default Isotopic Distribution” 

Table 5-4 of the TBD lists default specific activities to be used when only total uranium results 

are available. The report does not explain when this table should be used instead of Table 5-1, 

which gives percentages of uranium, neptunium, and plutonium activities at specific facilities, or 

Table 5-2, which presents default specific activities of uranium isotopes, as discussed above. 

This is a source of possible confusion for dose reconstructors. Not only do Tables 5-2 and 5-4 

employ different units—Bq/μg and nCi/g, respectively—the unit conversions for the uranium 

isotope are imprecise. The specific activities of the uranium isotopes in Table 5-4, when 

converted to Bq/g, are consistently about 3% higher than the default activities listed in Table 5-2. 

Although this is not a significant difference in terms of dose reconstruction, it is another example 

of the lack of accuracy and scientific rigor in the report. The table appears to be an amalgam of 

data from various sources—the report does not indicate how the values were calculated. 

As stated earlier in this review, this guidance ignores the information presented by BJC (2001, 

Table 2.4-1), which lists maximum concentrations of 99Tc, 237Np, and plutonium for 11 different 

operations at various facilities in the PGDP in various time periods. NIOSH should evaluate 

these data for use in the dose reconstructions of workers involved in these operations. 

We re-evaluated the default activity of each isotope listed in Table 5-4, as well as ones that were 

believed to be present at the PGDP but were not listed in this table. The discussion of each 

isotope is presented in order of atomic number.4

 The recommended changes in the values should 

be used in the generic dose reconstructions for workers, except for those known to have been 

exposed as a result of the operations described by BJC (2001), for whom the BJC data may lead 

to a more claimant-favorable dose reconstruction. 

5.4.2.4.1 Technetium-99 

The text cites a range of mass fractions of 99Tc of 0.041–7 ppm. This range actually represents 

two discrete estimates—the lower figure, cited by Smith (1984, Appendix 12) is the average 

concentration in French reactor tails, while the 7 ppm value is taken from DOE 2000. If the 

value of 7 ppm 99Tc in reactor tails was adopted as the concentration in airborne dust, the correct 

activity concentration would be 120 nCi/g U, not 0.12 nCi/g, as listed in Table 5-4. This is 

4

 The following review encompasses all isotopes relevant to dose reconstruction at the PGDP. It identifies 

isotopes whose default values are adequately characterized, as well as issues that need to be addressed. 
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another computational or editorial error. The activity listed in the table, therefore, understates 

the activity in the reactor tails by a factor of 1,000. However, concentrations of 99Tc in enriched 

reactor tails (ERT) averaged 16 ppm (uranium basis) (Smith 1984, Appendix 9), which is 

equivalent to about 274 nCi/g U. 

The concentrations in reactor tails do not necessarily reflect airborne concentrations. According 

to a discussion of TRU and technetium in DOE 2000; “However, these elements are much more 

hazardous than natural uranium and were concentrated by the cascade at certain specific 

locations, presenting increased hazards to certain workers.” According to PACE/Utah (2000, 

Appendix D), 99Tc concentrations in breathing zone air samples at ash handling locations ranged 

from 288 to 2,878 dpm/m3 β activity, vs. uranium α activities of 15–150 dpm/m3 . These ranges 

yield geometric means (GMs) of 910 and 47 dpm/m3, respectively, for a ratio of 99Tc to total 

uranium activity ≈19. According to PACE/Utah (2000, Table 7.9), the average fraction of 

uranium α activity in the dust at each of these two locations is 0.7 of the total α activity. Since 

the ash handled in these locations primarily consists of feed material, it is reasonable to assume 

the relative fractions of the uranium isotopes to be the same as those of natural uranium, with a 

total α activity of 686 nCi/g U. Based on these data, the specific activity of 99Tc in the dust is 

calculated to be about 19 μCi/g U (0.686 × 19.19 ÷ 0.7 = 18.8). This value constitutes a 

claimant-favorable default specific activity of 99Tc. 

5.4.2.4.2 Thorium-230 

The default specific activity of 230Th in Table 5-4 is equal to the highest value for this nuclide for 

the five locations listed in Table 5-3. However, PACE/Utah (2000, Table 7.9) lists two 

locations—ash receivers and pulverizer—with far higher fractions of 230Th in airborne dust. 

(The data tabulated by PACE/Utah for the C-400 converter salvage line is internally inconsistent 

and, therefore, is not being used in this comparison.) The average fraction of 230Th α activity in 

the dust at each of these two locations is 0.143, while the uranium accounts for 0.7 of the α

activity. Since the ash handled in these locations primarily consists of feed material, it is 

reasonable to assume the relative fractions of the uranium isotopes to be the same as in natural 

uranium, with a total α activity of 686 nCi/g U. The specific activity of 230Th is thus calculated 

to be 140 nCi/g U (686 × 0.143 ÷ 0.7 = 140), over 7 times higher than the value of 18.9 nCi/g U 

listed in Table 5-4. Since the default activity ratios could be used for workers at these locations, 

the higher value should be adapted for a conservative, claimant-favorable dose assessment. 

5.4.2.4.3 Uranium-234 

The default specific activity of 234U is listed in Table 5-4 as 702.0 nCi/g U. We believe the more 

conservative, claimant-favorable value of 0.0389 Bq/μg U listed in Table 5.4-3 should be 

adopted. This is equal to approximately 1,050 nCi/g U. 

5.4.2.4.4 Uranium-235 

The default specific activity of 235U is listed in Table 5-4 of the TBD as 43.9 nCi/g. This is in 

reasonable agreement with the value of 1.54 × 10-3 Bq/g (42 nCi/g) listed in Table 5.4-2. 
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5.4.2.4.5 Uranium-236 

The default specific activity of 236U is listed in Table 5-4 of the TBD as 0.93 nCi/g U. The 

highest enrichment of 236U in the PGDP cascade products were observed in May 1973, with an 

average concentration of 0.0625 wt% being measured in three cylinders withdrawn that month, 

which yields a value of 41.6 nCi/g U as a reasonable upper bound of 236U specific activity (see 

Table 5.4-2). If the dose reconstructor has no detailed information about the time and location of 

exposure, this conservative, claimant-favorable value should be adopted. 

5.4.2.4.6 Uranium-238 

The default specific activity of 238U, 337.5 nCi/g, listed in Table 5-4 of the TBD, is 

approximately that of this isotope in natural uranium. 

5.4.2.4.7 Neptunium-237 

As was the case with 230Th discussed above, the default specific activity of 237Np listed in 

Table 5-4 of the TBD is equal to the highest value for this nuclide for the five locations listed in 

Table 5-3. PACE/Utah (2000, Table 7.9) lists three locations—ash receivers, pulverizer, and 

converter maintenance—with higher fractions of 237Np in airborne dust. The average fraction of

237Np α activity in the dust in the first two locations is 0.007, while the uranium accounts for 

0.7 of the α activity. For converter maintenance, the activity fractions of 237Np and of uranium 

are both listed as 1, and the average air concentrations of these two nuclides are both listed as 

2.3E-11 μCi/cc. The same activity fractions and relative concentrations apply to the maximum 

values of these nuclides at that location. 

These activity fractions are obviously erroneous—the fractions of the two nuclides cannot both 

equal 1—which led us to question the absolute air concentrations as well. We confirmed the 

average 237Np concentration from data presented by Davis (1978), who reports activity 

concentrations in airborne dust during converter disassembly in Building C-409. The average

237Np concentration, derived from 18 samples (including both breathing zone and general area 

samples) is 50.3 dpm/m3, which is, in fact, equal to 2.3 × 10-11 μCi/mL. However, the average 

uranium concentration is equal to 7.8 × 10-11 μCi/mL. Uranium thus accounts for 76.4% of the α

activities, while 237Np constitutes 22%, the remainder being identified as 230Th and 239Pu. These 

uranium and neptunium values are almost exactly the same as those listed in Table 5-1 for 

Building C-409; similar values are listed for Building C-337. The specific activity of 237Np is 

therefore about 29% (0.22 ÷ .0724 = 0.288) of the combined specific activities of all the uranium 

isotopes. The total of the recommended specific activities of the uranium isotopes, as discussed 

above, is 1,463 nCi/g, yielding a specific activity of 237Np of 544 nCi/g, 100 times higher than 

the default activity of 5.4 nCi/g listed in Table 5-4. Since the default activity ratios could be 

used for workers at these locations, the higher value should be adapted for a conservative, 

claimant-favorable dose assessment. 
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5.4.2.4.8 Plutonium-238 

There is no mention of 238Pu in the TBD. The feed material for the PGDP cascades included 

Hanford reactor tails (Smith 1984). One source of information on plutonium isotope fractions of 

different materials produced at Hanford is PNNL 2004 (Chapter 8), which lists examples of 

mixtures of plutonium isotopes. Because 238Pu has a significantly shorter half-life than 239Pu 

(87.7 vs. 24,110 y), it has a correspondingly higher specific activity. The data listed by PNNL 

indicates that the 238Pu activity can exceed the combined activities of 239Pu and 240Pu in some 

plutonium fuel mixtures. It is not clear if 239Pu α activity listed in reference documents includes 

238Pu. Although we are not proposing a default specific activity for this isotope, we believe that 

this isotope should be accounted for in the PGDP dose assessments. We note that the internal 

dose coefficients for this isotope are similar to those for 239Pu, so that combining the activities of 

238Pu, 239Pu, and 240Pu would be an efficient and scientifically correct method of performing dose 

reconstruction. 

5.4.2.4.9 Plutonium-239 

As was the case for 230Th and 237Np, discussed above, the default specific activity of 239Pu listed 

in Table 5-4 of the TBD is equal to the highest value for this nuclide for the five locations listed 

in Table 5-3. However, as was the case for the other two nuclides, PACE/Utah (2000, Table 7.9) 

lists two locations—ash receivers and pulverizer—with far higher fractions of 239Pu in airborne 

dust. The average fraction of 239Pu α activity in the dust at each of these two locations is 0.15, 

while the uranium accounts for 0.7 of the α activity. Since the ash handled in these locations 

primarily consists of feed material, it is reasonable to assume the relative fractions of the 

uranium isotopes to be the same as natural uranium, with a total α activity of 686 nCi/g U. The 

specific activity of 239Pu is calculated to be 147 nCi/g U (686 × 0.15 ÷ 0.7 =147), more than 

twice the value of 67.5 nCi/g U listed in Table 5-4. Since the default activity ratios could be 

used for workers at these locations, the higher value should be adapted for a conservative, 

claimant-favorable dose assessment. 

Higher specific activities of 239Pu have been reported at PGDP. PACE/Utah (2000) reports the 

following: 

There was one documented case where a batch of material was sent to Fernald in 

June of 1980 that was found to have particularly high levels of plutonium. 

According to one document [8], the plutonium levels ranged from 67 to 7,757 ppb 

U. 

A concentration of 7,757 ppb 239Pu would correspond to a specific activity of 470 nCi/g. 

However, since this was the upper end of the range in one batch of material, it is unlikely that 

such a high value would represent the average specific activity in the airborne dust inhaled by 

any worker. 
 

 

 

 

 

 

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5.4.2.4.10 Plutonium-240 

No value for 240Pu is listed in Table 5-4 of the TBD. Smith 1984 states that the reported 239Pu 

activities included 240Pu, calculated as 239Pu. Since the dose coefficients for these two isotopes 

are virtually identical, this agglomeration does not affect dose reconstruction. However, this fact 

should be noted in the TBD. 

5.4.2.4.11 Plutonium-241/Americium-241 

The default specific activity for “241Pu/241Am” in Table 5-4 of the TBD is set to the same value 

as that of 239Pu. Since feed material for the PGDP cascades included Hanford reactor tails 

(Smith 1984), the specific activities of these nuclides can be estimated from the relative specific 

activities of Hanford reactor fuel mixtures, such as those listed by PNNL (2004, Chapter 8). 

According to PNNL, the 241Pu activity can be up to 144 times as great as the combined activities 

of 239Pu and 240Pu, while the 241Am activity can be up to four times as great. The default 

activities of these nuclides should be increased in light of these data. 

5.4.2.4.12 Other Fission Products 

Baker (1987) notes the presence of 106Ru, 95Zr, and 144Ce in the UO3 feed, and 106Ru and 95Zr in 

the ash, while Smith (1984, pg. 18) also cites the presence of 137Cs in the dust. Although these 

appear to be minor radioactive constituents, they need to be addressed in the TBD. 

5.4.2.4.13 Summary of Recommended Changes to Table 5-4 

The recommended changes to Table 5-4 of the TBD are listed in Table 5.4-3, below. We have 

not included fission products, other than 99Tc. 

Table 5.4-3. Recommended Changes to Default Specific Activities 

Radionuclide TBD SC&A 

nCi/g U Bq/μg U nCi/g U 

Tc-99 0.12 6.95E-01 18,800 

Th-230 18.9 5.18E-03 140 

U-234 702 3.89E-02 1052 

U-235 43.9 1.54E-03 42 

U-236 0.93 1.48E-03 40 

U-238 337.5 1.22E-02 329 

 Total U 1084.33 5.41E-02 1,463 

Np-237 5.4 2.01E-02 544 

Pu-239/240a 67.5 5.44E-03 147 

Pu-241b 67.5 7.83E-01 21,200 

Am-241b 67.5 2.18E-02 588 

Note: Fission product activities should be investigated and included if they
 

can make a significant contribution to worker doses.
 a

 Combined activity of both isotopes—see text.
 b

 Based on the recommended default activity of 239/240Pu and on the 

discussion in the text.
 
 

 

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5.4.2.5 Table 5-5: Facility-Specific Source Radionuclides with Solubility Type and 

Particle Size 

Table 5-5 of the TBD specifies the chemical compound, the lung clearance type (called 

“absorption type” in the table), and the particle size distribution for various radionuclides at a 

number of facilities at the PGDP, as well as default parameters listed at the end of the table. 

We have reviewed the literature cited in the footnotes to Table 5-5. One reference is listed as 

“PACE et al. (2000, Table 2.4-1).” This is an error—the table number does not correspond to 

any table in that document. 

The format of this table is needlessly long and cumbersome. Since many buildings have 

identical characteristics in terms of the radionuclides and their lung clearance types, the table 

would be much more usable if these buildings were listed together, rather than each building 

having a separate block of the table. Furthermore, the buildings are not listed in numerical order, 

nor in any other discernable order, requiring the reader to hunt through the entire table to locate a 

particular building. 

We will first discuss issues that apply to a number of buildings, following which we will discuss 

individual buildings, grouping together those having identical characteristics.5 

5.4.2.5.1 Particle Size 

The particle size in each case is specified as 5 μm AMAD. Such a particle-size assignment is not 

supported by the data, nor is it claimant favorable. The author cites a study that characterized the 

particle sizes of neptunium and plutonium isotopes as 3 μm in seven buildings at the PGDP, then 

dismisses these results in favor of a default of 5 μm recommended by ICRP 1994, which applies 

to a generic workplace exposure. Baker (1987, pg. 5) reports: “The UO2F2 fume had an AMAD 

of about 1 μm...” In discussing the exposure of cold trap and refrigeration systems operators, 

Baker states: “There was also chronic exposure to dust generated in other areas... (AMAD 

estimated at 4 μm)” (Baker 1987). Finally, Bruner (1960) cites that the dust particles at the 

PGDP are “about 0.5μ.” 

This decision to ignore site-specific data is contrary to both the intent and the letter of 

42 CFR 82. Furthermore, such a particle-size assignment is not claimant favorable. We have 

compared the 50-year doses to each organ from 3-μm and 5-μm AMAD particles of 230Th, the 

three natural uranium isotopes, 237Np, and 239Pu from all lung clearance types listed. Except for 

the extra-thoracic airways (in some instances), all organs would receive equal or higher doses 

from the 3-μm particles, the differences ranging from zero to 47%. Since particle sizes ranging 

from 1–10 μm AMAD are cited by Baker, the dose reconstructors should be instructed to use 

whatever particle size is the most claimant favorable in each case. 

5

 As in our discussion of Table 5-4 of the TBD, our critique of Table 5-5 is comprehensive and cites data that are 

adequately characterized, as well as those that are at issue. 
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5.4.2.5.2 Chemical Compounds—General Observations 

In a number of buildings, the compound is listed as UF6; however, upon contact with the water 

vapor in the air, UF6 is hydrolyzed according to the following reaction: 

UF6 + 2H2O = UO2F2 + 4HF 

Since UO2F2 has the same lung clearance type as UF6, this chemical reaction does not affect the 

internal dose corresponding to a given airborne concentration of uranium, but it should be noted 

in the TBD in the interest of scientific accuracy. 

The only specific chemical compounds listed for each building are compounds of uranium. The 

structure of the table gives the erroneous impression that these compounds apply to all the 

radionuclides, including isotopes of technetium, thorium, protactinium, neptunium, plutonium, 

and americium. For elements for which more than one lung clearance type is listed by ICRP 

(1994), the chemical compound needs to be listed in order to enable the correct assignment of a 

clearance type. If the compound is unknown, that should be stated in the table. Terms such as 

“ash handling” and “neptunium recovery” are improperly listed under the heading “Compound.” 

5.4.2.5.3 Uranium Progeny 

In each instance that 234mPa is included in the list of radionuclides, it is preceded by 230Th. This 

is clearly a propagation of the error at the beginning of Section 5.2, which we have noted earlier 

in this review. Protactinium-234m, which has a half-life of 1.17 minutes, would always be found 

together with its parent, 234Th. The intake of 234mPa makes a negligible contribution to internal 

dose; in fact, this nuclide is not even listed in ICRP Publication 68 (ICRP 1994). 

Thorium-230 is a long-lived (t1⁄2 = 75,380 y) daughter product of 234U that is found in 

conjunction with uranium that has aged for a number of years following its chemical separation. 

PACE/Utah (2000, Table 7.9) lists 230Th activities in Buildings C-410 and C-420, and in the cold 

trap area. Baker (1987) lists 230Th in the UO3 powder area, the C-420 green salt plant, the 

fluorination tower area, the cold trap and refrigeration operation, and Building C-410. Although 

Table 5-3 of the TBD includes 230Th activities in these locations but not in additional ones listed 

in PACE/Utah 2000, Table 7.9 (see discussion of Table 5-3 earlier in this review), this nuclide is 

not listed in Buildings 410 and 420 in Table 5-5. 

5.4.2.5.4 Transuranics 

Table 5-5 limits TRU to 237Np and 239Pu. However, as was discussed in the review of Table 5-4, 

the TRU in the Hanford reactor tails and other sources include 238Pu, 240Pu, and 241Am. These 

nuclides should be included with 237Np and 239Pu in Table 5-5, to be consistent with Table 5-4 

(which lists all but 238Pu), SAIC (1999, pg. 14), and Hill and Strom (1993, Table 16.2, Note “a”). 

Information about the occurrence of TRU nuclides at different facilities is presented by 

PACE/Utah (2000, Appendix D), which lists various radiological data. Air sampling, radiation 

survey, and urinalysis data indicates the presence of one or more TRU nuclides in the following 
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facilities: C-310, -315, -331, -333, -335, -337, -340, -360, -400, -409, -410, -420, -710, -720, 

-746, and -746B. In addition, Hill and Strom (1993, Table 16.2) list TRU nuclides at Buildings 

C-746A and “C-7460” (perhaps a typo for C-746Q). Finally, SAIC (1999, Table 1.7) lists TRU 

activities at C-333A, -337A, -404, -411, and -746Q. Since the dose assessments should always 

give the claimant the benefit of the doubt, the list of radionuclides in Table 5-5 should be all 

inclusive; therefore, the complete list of TRU nuclides should be included for each of these 

facilities. This requires expanding the radionuclide lists for C-310, -315, -337, -333A, -337A, 

-340, and -404. 

5.4.2.5.5 Technetium-99 

According to SAIC (1999, Table 1.7), 99Tc is found in the same locations as the TRU. (This is to 

be expected, since both products originate in recycled uranium). Thus, this nuclide should be 

added to the nuclide lists for C-315, -337, -333A, -337A, -340, -400, -404, and -409. 

Furthermore, DOE (2000, Appendix B) specifically cites high concentrations of 99Tc in process 

solutions in Bldg. C-400. 

5.4.2.5.6 Lung Clearance Types and Chemical Compounds 

The Lung Clearance Types, referred to as “Absorption Type” in Table 5-5, should be consistent 

with the chemical forms of each radionuclide. In case of uncertainty, the most claimantfavorable assumption should be adopted. It is not always clear which chemical form and/or 

clearance type the dose reconstructor should assign to each element. 

Technetium-99 

Technetium-99 at all facilities at the PGDP is assigned to Type F, the default for unspecified 

compounds (ICRP 1994, Annexe F). However, since technetium halides, as well as oxides and 

hydroxides, are Type M, that type should also be listed. According to DOE (2000, pg. 31), 

technetium formed a volatile compound of fluorine. (This was most likely TcF6, which boils at 

55.3oC, just below the sublimation point of UF6.) For 3 μm particles, the 50-year doses to the 

lungs and extra-thoracic airways are orders of magnitude higher from Type M; however, for 

some organs, the doses from Type M are up to 11% lower. Therefore, both types should be 

listed, allowing the dose reconstructor to select the more claimant-favorable type for a given 

organ. 

Thorium-230 

Thorium-230 is assigned solely to Type S. This assignment is not scientifically correct. 

According to ICRP (1994), Type S is assigned to thorium oxides and hydroxides, while all other 

compounds are assigned to Type M. Thorium fluorides (e.g., ThF4) at the feed plant are 

mentioned by Baker (n/d, pg. 389); thus, Type M thorium should also be considered. 

Furthermore, the assignment to Type S may not be claimant favorable; for a given intake, 

Type M delivers doses to organs other than the lung and the extra-thoracic airways that are 9–
 

10 times higher than those from Type S. However, in cases where 230Th was detected in the 

urine, the calculated dose based on the concentration in the urine would be higher if the thorium
 

were assigned to Type S. Therefore, both types should be listed, allowing the dose reconstructor 

to select the more claimant-favorable type for a given organ. 
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Uranium 

Buildings C-310, C-315, C-331, C-333, C-335. Uranium isotopes at Buildings C-310, C-315, 

C-331, C-333, C-335, which house the cascade, are assigned to Type F. The uranium was in the 

form of UF6 which, upon release to the atmosphere, reacted with moist air to form UO2F2, as 

discussed previously. Both of these compounds are Type F, so the assignment is appropriate. It 

is not clear why the uranium in Buildings C-333A and C-337A is assigned to Type M as well as 

F. The only compound listed in Table 5-5 is UO2F2, which is Type F. If other compounds are 

believed to be present, this should be explained in the TBD. 

Building C-340. Uranium isotopes in Building C-340 are restricted to Types M or S. However, 

the list of compounds includes UF6, which is Type F. Type F should therefore be included as an 

option for use by dose reconstructors for workers exposed at this facility. 

Building C-360. Uranium isotopes in Building C-360 are assigned to Types F, M, or S. Since 

the building appears to have been used for transfer and sampling of UF6 cylinders, it is not clear 

why Type M or S compounds would be found there. In the interest of equity in dose 

reconstruction, the same logic in assigning lung clearance types should be used for all facilities. 

Building C-400. Uranium isotopes in Building C-400 are assigned to Types F, M, or S; 

however, the only compounds listed are UF6 and UO2F2. According to DOE (2000, 

Appendix B), releases of UF4 (Type M) and U3O8 (Type S) also occurred in this building. These 

compounds should be listed in Table 5-5 to document the assignment of lung clearance types. 

Facility C-404. Uranium isotopes in Facility C-404 are assigned to Types F or M. However, 

since the wastes disposed of in this area included incinerator ash, the presence of UO2 is likely—
 

Type S should also be included. For the sake of completeness, the list of compounds should 

include UO2F2 if there is reason to believe that this compound (or the dry form, UF6) was 

released there. If not, there may be no justification for including Type F as an option. 

Building C-405. Uranium isotopes in Building C-405 are assigned to Types F, M, or S. 

Building C-409. Building C-409 is described as the decontamination building, whereas 

PACE/Utah (2000) and Turpin (2006) call it the stabilization building. The terminology should 

be corrected. Uranium isotopes in Building C-409 are assigned to Types F or M; however, the 

only compounds listed are UF6 and UO2F2. These are also the only uranium compounds listed 

by DOE (2000, Appendix B). Unless there is evidence for UF4 or other Type M compounds at 

this facility, Type M should not be included in Table 5-5. 

Buildings C-410 and C-420. Buildings C-410 and C-420 housed the feed plant. Thus, uranium
 

compounds include UO2, UO3, U3O8, UF4, and UF6 (which forms UO2F2 when released). 

Table 5-5 includes “UF” (a nonexistent compound) in the list for both buildings and omits U3O8. 

It also omits UF6 (or UO2F2) at C-420. The assignment of the uranium isotopes to Types F, M, 

or S is correct, given the possible compounds. 

Building C-710. Building C-710 housed the analytical laboratories, which performed 

neptunium and uranium recovery operations. Absent specific information on the uranium
 
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compounds handled in this building, the claimant-favorable assumption should be that all three 

lung clearance types of uranium—F, M, and S—could have been present. Representative 

compounds of each clearance type (e.g., U2O2F2, UF4, and UO2) should be listed in the 

“Compound” column. 

Building C-720. Building C-720 housed various maintenance operations, including converter 

and compressor disassembly and flange grinding. Workers were primarily exposed to UF6 and 

UO2F2, but also to UF4 and oxides such as U3O8 or UO2. These compounds represent all three 

lung clearance types—F, M, and S. Table 5-5 lists these three types for 235U and 238U, but 

erroneously omits Type S for 234U. Uranium compounds representing all three classes should be 

listed in the “Compound” column. 

C-746 Facilities. DOE (2000) refers to a number of C-746 facilities: C-746C, C-746E, C-746F, 

C-746K, C-746R, C-746S, and C-746T, in addition to C-746A, C-746B, and C-746Q, which are 

listed in Table 5-5. It would be simplest to list “C-746” as a general category that includes all 

these facilities. Given the diversity of operations, all three clearance types and their 

representative compounds should be included. 

Neptunium 

All neptunium compounds are assigned to Type M (ICRP 1994, Annexe F). However, for a 

number of facilities, Table 5-5 erroneously includes Type F, and sometimes also Type S, for 

237Np. Such assignment, for which no dose coefficients have been developed, is a possible 

source of confusion for dose reconstructors. 

Plutonium 

All plutonium isotopes at all facilities are assigned to Type S in Table 5-5. This assignment is 

not scientifically correct. According to ICRP 1994, Type S is assigned to insoluble plutonium 

oxides, while Type M applies to all other compounds. According to DOE (2000), “... most of 

the plutonium and technetium was volatilized to the cascade...” This indicates that plutonium 

was most likely present as a fluoride rather than an oxide; PuF6 melts at 52oC, while PuO2 has a 

melting point of 2400oC. Furthermore, such an assignment may not be claimant favorable—for a 

given intake, Type M delivers doses to organs other than the lung and the extra-thoracic airways 

that are 9–10 times higher than those from Type S. However, in cases where plutonium was 

detected in the urine, the calculated dose based on the concentration in the urine would be higher 

if the plutonium were assigned to Type S. Therefore, both Type M and Type S should be listed, 

except for facilities where only one or the other class of compounds was present. For instance, it 

is likely that only Type M would be present in the cascades, since the exposure would be to 

releases of PuF6 in the UF6 gas. 
 

 

 

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5.4.3 Section 5.3: “In-Vitro Measurement Methods” 

5.4.3.1 Calculation of Intakes 

5.4.3.1.1 Frequency of Sample Collection 

Table 5-6 of the TBD lists the frequency of in-vitro measurements at various facilities at the 

PGDP during various time periods. The frequencies range from once every 4 weeks to once a 

year. The last row of the table lists a default frequency of once every 4 weeks. Since the interval 

between measurements could have been as long as 1 year, this default assumption is not claimant 

favorable. The longer the interval over which the intakes occurred, be they chronic or acute, the 

lower the urine concentration for a given total intake. Conversely, for a given urine 

concentration, the longer the elapsed time (for acute intakes) or exposure duration (for chronic), 

the greater the derived intake. 

To illustrate this effect, we have calculated the acute intake that would result in the excretion of 

1 Bq in 24 hours, assuming the inhalation of Type F, M, or S uranium with an AMAD of 5 μm. 

The results are shown in Table 5.4-4 of this review. 

Table 5.4-4. Acute Intake of Uranium (Bq) That Would Result in Excretion of 1 Bq/d 

 Time After Lung Clearance Type 

Intake (mos) F M S 

1 1.47E+03 3.77E+03 1.30E+05 

2 4.41E+03 5.99E+03 1.93E+05 

4 1.39E+04 1.02E+04 2.63E+05 

As shown in Table 5.4-4, in the case of Type F uranium, assuming that an intake occurred 

1 month before the urine sample was collected, when in fact it occurred 4 months ago, would 

result in a 10-fold error in estimating the intake. 

5.4.3.1.2 Converting Spot Samples to 24-hour Samples 

Equations 5-1 to 5-3 of the TBD present methods of converting spot samples to 24-hour samples. 

We have several comments on the formal presentation of these equations, as well as on their 

content. The first observation is that the physical units should be included in a description of 

symbols that precedes or follows the equation, not in the equation itself. The equations as 

written are incorrect and confusing. We recommend that Equation 5-1 be written as follows: 

Ccor	 = corrected activity concentration in urine sample (pCi/L) 

Cm	 = measured activity concentration (pCi/L) 

SGr	 = reference specific gravity of urine (dimensionless) 

= 1.024 
 

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SGm = measured specific gravity of sample 

Equation 5-2 in the TBD incorrectly indicates that the corrected concentration is multiplied by 

1.4 and divided by 24. We recommend that the equation be written as follows: 

Au	 = activity excreted in 24 h (pCi) 

V = excretion rate for Reference Man (ICRP 2002) 

= 1.6 L/d (adult man) 

= 1.2 L/d (adult woman) 

Similarly, Equation 5-3 should be written: Cm = measured activity concentration (pCi/L) 

In the line immediately above Equation 5-3, the text “by assuming a reference value V of 1.4 and 

volume L of 1.0 L” should be deleted. The text as it stands is incorrect—“L” is the symbol for 

liters, not some specified volume—and is no longer needed if the above corrections are adopted. 

Furthermore, Equation 5-2 in the TBD is based on an excretion rate of 1.4 L/d of urine. 

However, this value is based on an earlier reference value for an adult man. The current ICRP 

(2002) reference values are 1.6 L/d for an adult man and 1.2 L/d for an adult woman. Thus, the 

value of 1.4 L/d is not scientifically valid for either male or female workers. Furthermore, it is 

not claimant favorable for males. 

5.4.3.2 Minimum Detectable Concentrations 

Table 5-7 of the TBD lists minimum detectable concentrations (MDCs) for total uranium and for 

various radionuclides, using different analytical techniques during different time periods at the 

PGDP. The column headed “Measurement type” lists analytical techniques in some cases, and 

simply “ORNL” or “contractor” in others. The types of measurements performed by ORNL and 

the unspecified contractor need to be presented, or an explanation given as to why this 

information is not available. Without any information on the type of analysis, it is not possible 

to independently assess the MDCs for these measurements. The column headings “MDC 

(mg/L)” and “Recall level (μg/L)” are confusing, since not all values are in these units. A 

footnote indicating that these are the units “unless otherwise noted” would help to clarify this 

issue. 

In many cases, we could not verify the listed MDCs. For example, the MDC for total uranium 

by in-house fluorimetry is listed as 10 μg/L. None of documents cited as data sources for this 

table list that value; both PACE/Utah (2000) and SAIC (1999) cite typical MDCs of 5 μg/L. The 

MDC for natural uranium assayed at ORNL from 1999 to the present is in error; the source 

document (SAIC 1999) lists the MDC at 0.06 μg/sample, while Table 5-7 lists it as 0.06 mg 

(60 μg), a 1000-fold discrepancy. The latter value is clearly inconsistent with the value of 5 μg/L 

for kinetic phosphorescence analysis (KPA), used in 1977–1982. Table 5-7 lists a default 

detection level of 0.27 pCi/L for urinalysis of individual isotopes of actinide elements (Th, U, 

Pu, and Am). The TBD cites ICRP Publication 54 (Annals of the ICRP Vol. 19 No. 1–3) , 
 

 

 

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Individual Monitoring for Intakes of Radionuclides by Workers: Design and Interpretation 

(Oxford: Pergamon Press) as the source of these data, which are based on alpha spectroscopy.6 

However, Table 5-7 lists higher levels—1 pCi/L for Pu isotopes and 0.3 pCi/L for uranium 

isotopes. If the dose reconstructor does not know the analytical method involved, this default 

assumption is neither claimant favorable nor scientifically correct. 

Given the importance of the MDC in assigning doses in cases of low readings or non-detection, 

greater attention should be given to this topic to make certain that the values listed result in 

claimant-favorable dose assessments. 

There is no discussion of the uncertainties in the determination of uranium in urine samples for 

fluorimetry and KPA. The uncertainty of the fluorimetry methodology can be very high; it is 

strongly dependent on the quenching of the sample. For other radionuclides, there is no 

description of the technique used for their measurements and the accompanying uncertainties. 

5.4.3.3 Day of Sample Collection 

Section 5.3.3 of the TBD states the following: 

The practice of offsite collection of samples that takes place 24 to 48 hr after 

leaving the plant not only minimizes the possibility of sample crosscontamination, but it ensures that samples are collected after the transfer of the 

rapid clearance component. Some PGDP employees were asked to collect 

samples after 1 or 2 days off from work; if so, that collection instruction was 

sometimes noted on the analytical record. 

The TBD fails to note that this practice would lead to a lowering of the calculated intakes, nor 

does it instruct dose reconstructors to be alert to any cases for which urine samples were 

collected after the worker was off work for any period of time. Appropriate adjustments to the 

calculated intake should be made to compensate for the lowered concentration in the urine 

following an absence from work. 

Table 5.4-5 of the present review illustrates the effect of the day of sample collection on the 

calculated intake. In this example, we have assumed a chronic intake by inhalation of Type F 

uranium with an AMAD of 5 μm by a worker who is exposed 8 hours per day, 5 days per week 

(Monday through Friday). We have calculated the intakes that would result in a rate of excretion 

of 1 Bq/d on each successive day of the week during the 4th week of the assessment period. 

6 We also note that ICRP Publication 54 has been replaced by ICRP Publication 78: Individual Monitoring for 

Internal Exposure of Workers - Annals of the ICRP Volume 27/3-4, Replacement of ICRP Publication 54, 1998 (Oxford: 

Pergamon Press). 
 

 

 

 

 

 

 

 

 

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Table 5.4-5. Intakes Derived from Samples Collected on Different Days of the Week 

Exposure Duration IRF: Intake Time of Sample (d) 24-hour Urinea (Bq/d) 

21 Monday morning 0.0402 24.9 

22 Tuesday 0.218 4.59 

23 Wednesday 0.223 4.48 

24 Thursday 0.226 4.42 

25 Friday 0.228 4.39 

26 Saturday 0.230 4.35 

27 Sunday 0.0504 19.8 

Note: Each sample assumed to correspond to an excretion of 1 Bq/d
 a Intake retention fraction
 

Table 5.4-5 shows significant differences in the calculated intakes for different days of sample 

collection. If the sample was collected on Wednesday, the calculated chronic intake during the 

previous 31⁄2 weeks would be about 4.5 Bq/d. However, if the sample was known to be collected 

on Monday morning, prior to starting work, the calculated intake during the previous 3 weeks 

would be about 25 Bq/d, over 5-times higher. The TBD should provide claimant-favorable 

guidance for dose reconstruction of cases where the date of sample collection is not recorded. 

This guidance becomes especially relevant in the case of recall samples, which were normally 

collected on Monday mornings (PACE/Utah 2000, pg. 38). Instead, the TBD states the 

following: 

... contamination of samples from the worker’s hands or clothing cannot be ruled 

out as a contributor to any given result. If a second analysis was performed and 

if that result was negative, it is reasonable to assume the first result was a false 

positive due to sample contamination or laboratory error. 

Such guidance violates 42 CFR 82, who states that in cases of uncertainty, the dose 

reconstruction should use assumptions that are claimant favorable. The TBD should instruct 

dose reconstructors to consider all results, even if a later analysis had a lower or non-detectable 

concentration. To do otherwise would create a bias against the claimant. 

5.4.3.4 Dietary Intake of Uranium 

The TBD gives confusing guidance regarding the possible dietary intakes of uranium by 

monitored workers: 

Because studies of the average daily uranium excretion on Paducah residents do 

not appear to have been performed, it is not possible to make corrections for the 

contribution of nonoccupational intakes of uranium to a given urine sample 

result. However, to put a given result into perspective, a nominal daily (24 hr) 

urinary excretion rate for uranium of 0.43 μg (environmental decision level at 

95% confidence) can be used... No correction for environmental levels of 
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uranium is required for samples analyzed by fluorimetry or KPA because the 

MDC is larger than the correction. [Emphasis added.] 

It is not clear if the TBD intends to give dose reconstructors the option of subtracting 0.43 μg 

from the 24-hr urine sample analyzed by techniques other than fluorimetry or KPA. Such 

guidance would be neither scientifically correct nor claimant favorable. First, the value cited by 

SAIC (1999) (the TBD refers to this document as BJC) applies to residents of the Oak Ridge, 

Tennessee, area, not of the PGDP. Second, it is an upper limit (95th percentile). Subtracting 

such a high-end value from the urinalysis result would result in a bias against the claimant. The 

TBD should state unambiguously that the results should not be corrected for any assumed 

contributions of uranium in the environment. 

5.4.3.5 Table 5-9: “In-Vitro Record Codes” 

Table 5-9 of the TBD lists the record codes for urine bioassays. First, we note that one of the 

sources for these data is listed as “PGDP 2003b.” That document, Uranium in Urine Analysis 

(n/d), contains no information on record codes. The author most likely meant to cite “Urinalysis 

Program” (n/d) which does, in fact, contain such data. 

For Form WCP-455, the codes for “Reason for Visit” contain one ambiguity; code 33 is 

identified both as “industrial health recheck” and “termination samples.” This ambiguity should 

be resolved. Furthermore, the midnight shift is identified as “O,” whereas on form WCP-885, 

the shifts are indicated by the letters A–D. This is a possible error that should be resolved. The 

cited documents do not provide any information on the coding of Form WCP-455; therefore, we 

could not verify the accuracy of these data. 

Urinalysis Program (n/d) presents instructions for coding Form WCP-885. We have several 

comments about the presentation of the codes for this form in Table 5-9 of the TBD. First, the 

code for “shifts” refers to four shifts, identified as A, B, C, or D. Table 5-9 erroneously 

identifies the third shift as “O.” Next, the frequency codes are erroneously based on an example 

of how to fill out this form, presented in Urinalysis Program (n/d). The correct interpretation of 

the frequency codes is: 1 = 1 per month, 2 = 1 per every 12 months. The identity of the analyte 

is indicated elsewhere on the form and is not relevant to this code. 

Form UCN-5242 is stated to have the same column identifiers and codes as “NCP-455.” It is 

most likely that the correct designation is “WCP-455.” 

Finally, we note that units are listed for breath and fecal analyses. The only other mention of 

fecal analyses in the TBD is a footnote that indicates that routine fecal analyses were not 

performed at the PGDP. According to PACE/Utah (2000), fecal bioassays were performed on 

16 workers. Should any of these workers or their coworkers become claimants, such data could 

be useful in performing dose reconstructions. Information on the fecal analysis methodology, 

the MDCs, and guidance on the use of these data in dose reconstruction should be provided in 

the TBD. 
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No information is presented on breath analyses. If such studies were performed, information 

similar to that needed for fecal analyses should be provided. 

5.4.4 Section 5.4: “In-Vivo Measurement Methods” 

Table 5-10 presents the MDAs for the measurement of various radionuclides by whole-body or 

chest counting. The sources cited, however, do not present the values listed, with the sole 

exception of 237Np for the 1968–1991 period. Some of the values are inconsistent. For example, 

2% enriched uranium has about 3 times the concentration of 235U as natural uranium, yet the 

MDA for “total uranium” is 40 times higher (4 mg vs. 100 μg) than 2% enriched uranium. The 

235U in natural uranium would permit the detection of smaller quantities of uranium. The same 

MDA (4 mg) is listed for 238U. Even the depleted uranium tails at the PGDP contained about 

0.27% 235U on a mass basis, which would be sufficient to permit the detection of less than the 

stated MDA. Uranium-238 can be detected by the radiation emitted by its daughter product, 

234Th, which has a 24-day half-life and an MDA of 3 nCi (SAIC 1999, Table 3.2). Assuming the 

two nuclides remain in secular equilibrium in the body, this corresponds to 10 mg of 238U, and 

approximately the same quantity of natural uranium. These MDA values should be verified and 

documented to enable verification by an independent reviewer. 

The MDA for in-vivo counting at Y-12 prior to 1967 appears to be overstated. Bruner (1960) 

reports a lower limit of 7 nCi, which is 2 orders of magnitude lower than the value of 0.5 μCi for 

the 1960–1967 period. 

Table 5-11 lists in-vivo record codes. Of the three references cited, “PGDP 2003a” (Urinalysis 

Program n/d) is clearly inapplicable. There is no available documentation for the record codes of 

the “In-Vivo Radiation Monitoring Report;” consequently, these codes could not be verified. 

5.4.4.1 Addressing Interferences and Uncertainties 

According to Section 5.4.3 of the TBD: 

For in vivo measurements, contamination could have occurred as external to the 

body or, in the case of chest counting, as external to the lung. If a follow-up in 

vivo count (the same day or within a few days) showed a dramatic decrease in 

activity or no detectable activity, then external contamination should be assumed. 

We agree that if a second in-vivo count performed shortly after the first (i.e., within 1–2 hours) 

showed a markedly reduced count rate, it is reasonable to assume that the worker underwent 

external decontamination following the first measurement, and that the second count was more 

indicative of the internal body burden of a given radionuclide. We do not agree, however, that a 

count taken several days later should be substituted for the earlier measurement. The first count 

could have detected activities in the lungs that were later transported to other organs. If the 

measurements were chest counts, then the counting efficiency of radioactive contaminants in 

other organs would be lower, both because the detector geometry was optimized for the lungs, 

and because abdominal organs, for instance, are shielded by a greater mass thickness of tissue, 

which absorbs more of the emitted photons. Therefore, it is neither correct nor claimant 
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favorable to discard the earlier counts on the unverified grounds that they may have been due to 

external contamination. 

The TBD gives no instruction to the dose reconstructor on estimating uncertainties in the in-vivo 

measurements, except to state, “For results near or at the reporting levels, dose reconstructions 

should apply the prescribed standard deviation of 0.3 times the MDA or reporting level...” 

While we do not disagree with that recommendation, clearer guidance is needed for higher 

levels. The uncertainties for in-vivo bioassay include the uncertainty in the counting statistics, 

including the background that is subtracted from the count rate, the uncertainty of translating a 

count rate from a given worker into a body burden, given the differences in morphology among 

different individuals, and the variation in the naturally occurring radionuclides in the human 

body, primarily 40K, that contribute to the background. 

During site interviews, it became apparent that only a subset of workers from a particular group 

was monitored, indicating cohort monitoring for in vivo counts. This leads to additional 

uncertainty in the use of in vivo counting and deserves further evaluation. 

The TBD instructs dose reconstructors to disregard any measurements of 137Cs, on the grounds 

that there was no occupational exposure to this radionuclide at the PGDP. We disagree with this 

conclusion, since this nuclide may have been present in reactor tails processed at the PGDP, and 

was, in fact, detected in cascade dust (Smith 1984, pg. 18). Therefore, although in-vivo 

measurements of 137Cs may be due to atmospheric deposition from worldwide testing of nuclear 

weapons, such measurements can be disregarded only if there were adequate measurements on 

individuals in the PGDP areas who had not had any occupational exposures to radioactive 

materials, and who could therefore serve as controls. Absent such data, the 137Cs measurements 

should be considered evidence of occupational exposure and included in the dose reconstruction. 

5.4.4.2 “Significan Incidents wi t th Internal Dose Potential”

Table 5-12 of the TBD is entitled “Input Parameters for Significant Incidents and Events.” In 

fact, this table presents a partial chronology of activities at the PGDP. Some of the entries, such 

as “guard patrolling,” have a “Low Potential for Increased Radiation Exposure” (PACE/Utah 

2000, Table 6.1) and do not belong in this table. Furthermore, there is no direction on how these 

and other events listed in Table 5-12 are to be used in dose reconstruction. 

Significant information that could be useful to dose reconstructors is not included in this section 

of the TBD. This information includes statements by workers that urine specimens were 

collected within 30 minutes of an incident or accident with a potential for elevated exposure. 

Such a short time period does not allow for equilibrium between the inhaled activity and the 

concentration in the urine, most likely resulting in a false negative. Follow-up samples were 

collected from workers who did show elevated levels of radioactive materials in the urine. More 

important are the statements of former workers that the bioassays performed following such 

incidents were not always recorded in the database. Thus, the doses from such incidents may not 

be in the worker exposure records. 
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5.5 OCCUPATIONAL EXTERNAL DOSE 

Statement of Purpose 

SC&A’s review of ORAUT-TKBS-0019-6, Rev. 01 (March 29, 2005) is intended to assess the 

technical merit as well as the completeness of guidance provided for the reconstruction of 

external doses associated with photon, electron, and neutron radiation environments to which 

workers may have been exposed at PGDP. 

For practical reasons, our review and citation of issues will parallel the sequence of information 

and data as presented in the TBD. 

Relevant Background Information 

The PGDP began operation in 1952. The principal process of PGDP involves the enrichment of 

feed material in the form of UF6 gas to about 2.5% U-235. Originally, most UF6 came from 

depleted tails produced during normal diffusion operations at PGDP, ORNL, and Portsmouth. 

From 1953 to 1964, some feed material (about 17%) was recycled uranium obtained from spent 

reactor fuel shipped from Hanford and SRS. In brief, feed material consisted of depleted, 

natural, and recycled uranium. 

The Site Description TBD (Turpin 2006) states the following: 

. . . an estimated 2500 to 4000 workers worked in areas with ‘moderate’ to ‘high’ 

potential for increased internal and external radiation exposures. This is based 

on a relative ranking of the potential of radiation exposures at PGDP. These 

areas included the feed plant (Buildings C-410 and -420) with operators and 

mechanics receiving the highest doses, respectively, decontamination building 

(C-400) with decontamination workers receiving the highest doses, and the 

cascade buildings (C-331, C-333, C-335, and C-337) with operators receiving the 

highest doses. In addition, workers in Buildings C-340 and C-720 had the 

potential for increased external radiation exposures. Average doses per 

department are listed in Table 2-1 . . . [and] Average and maximum doses are 

listed in Table 2-2. 

. . . Approximately 10% of the 2500 to 4000 workers had a potential for higher 

than average radiation exposures. [Emphasis added.] 

Table 2-1 cited above and Turner’s Table 6-5, which incorporates the data from Turpin’s 

Table 2-2, are reproduced herein as Table 5.5-1 and Table 5.5-2, and will be referenced in 

discussions below. In support of data presented in these tables, Turner (2005, Section 6.2) states 

the following: 

From startup until July 1960, PGDP issued dosimeters to a limited number of 

individuals (PACE/Utah 2000). This population of monitored individuals 

represents those with the highest exposure potential. After July 1960, PGDP 
 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

 

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routine practices required the assignment of dosimeters to all workers who 

entered a controlled radiation area... [BJC 2001] [Emphasis added.] 

Table 5.5-1. Average External Cumulative Penetrating Dose (deep dose equivalent) and 

Number of Workers Assigned to Each Department During 1953–1988 

Dept. No. Job Description 

Average 

Cumulative Dose 

(mrem) 

Number of 

Workers 

5751 Feed plant operators 3,814 185 

5760 Decontamination 2,788 116 

5034 Feed plant mechanics 2,587 99 

5015 Unknown 2,025 17 

5676 Unknown 861 14 

5730 Cascade operators 627 578 

5785 Chemical operators 595 113 

5075 Instrumentation 538 245 

5020 Unknown 481 17 

5008 Transport pool 371 33 

5002 Process maintenance 364 578 

5108 Environmental control 338 48 

5268 Unknown 316 236 

5077 Electricians 298 318 

5005 Material termination mgr. 295 90 

5772 PEMU decontamination 253 22 

5759 Unknown 220 4 

5049 Unknown 182 12 

5725 Unknown 175 20 

5044 Mech. Inspection 170 113 

5021 Plant services 147 486 

5770 Converter test 145 23 

5035 Feed plant mechanics 143 160 

5019 Unknown 142 13 

5740 Nitrogen plant 142 22 

5646 Metals building 132 95 

5674 Unknown 129 8 

5048 Fabrication shops 127 667 

5023 Unknown 115 24 

5675 Unknown 114 7 

5743 Steam plant 111 61 

5027 Unknown 110 282 

Total 4,706 
 

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Table 5.5-2. Average Recorded External Radiation Doses (deep dose equivalent) per 

Worker per Year from 1953 to 1988 including Maximum Exposures Recorded for any 

Single Worker for that Year 

Year Number of 

Workers 

Average Dose, 

GM (rem) 

Maximum Dose 

(rem) GSD (rem) 

1953 223 0.1398 0.820 2.14 

1954 284 0.2835 1.580 2.09 

1955 417 0.2419 2.500 2.72 

1956 471 0.3586 4.700 3.02 

1957 669 0.2517 3.190 2.97 

1958 661 0.1853 3.630 3.59 

1959 570 0.2015 2.360 2.88 

1960 526 0.2011 2.510 2.95 

1961 1,690 0.1770 2.530 3.13 

1962 1,479 0.1495 2.980 3.01 

1963 1,311 0.1441 3.040 3.70 

1964 1,289 0.0734 1.860 4.00 

1965 1,128 0.0341 1.610 5.23 

1966 1,138 0.0371 1.470 5.19 

1967 1,143 0.0498 1.120 3.80 

1968 1,241 0.0618 1.400 3.82 

1969 1,270 0.0733 1.970 4.11 

1970 1.273 0.0417 0.840 3.63 

1971 1,254 0.0624 1.380 3.78 

1972 1,288 0.0589 1.760 4.30 

1973 1,404 0.0530 1.830 4.57 

1974 1,624 0.0265 1.030 4.81 

1975 2,013 0.0501 1.049 3.69 

1976 2,426 0.0351 1.224 4.59 

1977 2,643 0.0232 0.742 4.42 

1978 2,613 0.0399 0.359 2.57 

1979 2,487 0.0082 0.364 5.09 

1980 2,308 0.0182 0.344 3.53 

1981 1,840 0.0076 0.420 5.60 

1982 1,617 0.0065 0.350 5.53 

1983 1,452 0.0067 0.340 5.39 

1984 1,434 0.0092 0.420 5.15 

1985 1,365 0.0061 0.350 5.69 

1986 1,244 0.0096 0.490 5.41 

1987 1,275 0.0080 0.470 5.74 

1988 1,359 0.0065 0.720 7.54 

Presented below are summary descriptions of issues that conflict with basic assumptions stated 

in the TBD, and which may impact the credibility/claimant favorability of dose reconstructions. 
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5.5.1 Issue 1: Questionable Assumptions for the Assessment of Recorded Skin Dose 

Section 6.3.2.1 of the TBD states the following: 

...The nonpentrating response of the two-element dosimeter was calculated as the 

difference between the “unshielded” and “shielded” portions of the film based on 

a uranium calibration. The two-element dosimeter workplace nonpenetrating 

(i.e., beta or shallow) dose response based on the uranium calibration should 

adequately represent HP (0.07) or at least be claimant-favorable because of the 

significant over-response of the unshielded portion of the film to any lower energy 

photons that could have been present... [Emphasis added.] 

The above quotation references a “two-element” dosimeter, yet Table 6-1 of the TBD only 

identifies four-element film dosimeters in use between 1953 and 1980. This “technical” 

discrepancy has significant implications when viewed in context with the statement that the 

resultant beta dose is likely to be claimant-favorable “. . . because of the significant overresponse of the unshielded portion of the film to any lower energy photons that have been 

present.” [Emphasis added.] 

Inspection of Figure 6-1 in the TBD shows that, while this statement may apply to the twoelement film, it appears not to apply to the multi-element film. 

The method for deriving the nonpenetrating response as “. . . the difference between the 

unshielded and shielded portions of the film” has been questioned by the National Research 

Council (1989), which stated the following: 

Pages 39-41: 

When a film badge with only a lead filter and an open area is exposed to unknown 

mixtures of beta and photon energies, it is not possible to determine contributions 

from each component to NOD [net optical density] in the film open area. 

. . . The final attempt to evaluate and report beta exposure with film badges 

during atmospheric testing was . . . during Operation PLUMBBOB in 1957. . . 

This badge has four filter areas: lead-tin laminate, open window, copper, and 

aluminum. This combination was thought to be capable of providing beta 

exposure, but the analytical procedures used were faulty . . . As stated previously, 

the function of NOD versus exposure is not linear, and NOD from a film must by 

converted to exposure with a common calibration curve because an increment of 

NOD can represent a different amount of exposure at different locations on a 

calibration curve. . . Thus, beta particle monitoring with personnel badges was 

not successful during atmospheric nuclear testing series. [Emphasis added.] 
 

 

 

 

 

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Section 6.3.3.1 (page 11) of the TBD provides the following statements: 

For a number of years, ORNL used uranium beta as well as Ra-226 gamma 

calibration curves to interpret film densities... The ratio of beta-to-gamma 

responses was tested in several ways. Films wrapped in a 7 mg/cm2

 absorber 

were placed in contact with a slab of natural uranium... [Emphasis added.] 

However, it was also acknowledged that because the actual dosimeter worn by individuals: 

. . . had a minimum absorber thickness of 80 mg/cm2

 between the film and the 

source, the effective beta energy is needed to interpret the film density in terms of 

HP (0.07). The radiation was routinely treated as 1.7 MeV beta particles from 

uranium, which are about 40% absorbed in 80 mg/cm2 . The determination of 

beta dose was thus specific to uranium [U-238]. 

Section 6.4.2 of the TBD ironically concludes the following: 

The early film dosimeters were calibrated to uranium for nonpentrating radiation. 

No numerical adjustment of recorded shallow doses is recommended. [Emphasis 

added.] 

The high-energy beta of 2.29 MeV is emitted from Pa-234, which is the short-lived daughter of 

U-238. However, this simple “calibration” factor ignores the following lower-energy betas from 

short-lived daughters of U-238 and U-235, and from Tc-99 that would be completely absorbed 

by the 80 mg/cm2 absorber: 

Beta Max (MeV) Yield 

Th-234 0.103 21% 

0.193 79% 

Th-231 0.206 13% 

0.287 12% 

0.288 37% 

0.305 35% 

Tc-99 0.294 100% 

While all of these betas would contribute to the 7 mg/cm2

 skin dose, none of these betas would 

have penetrated the 80 mg/cm2 absorber and registered as a “dose” on the open window (see 

Exhibit #1, which defines the range of betas in an absorbing medium). Inspection of Exhibit #1 

shows that a beta particle with energy >300 keV is required to penetrate an absorber of 

80 mg/cm2 . The magnitude of this deficiency is substantial and variable, since it is affected by 

the degree of U-235 enrichment and the contamination level of Tc-99. 

SC&A concludes that the combination of these factors is likely to have resulted in skin dose 

estimates that are inaccurate, low, and not claimant favorable. 
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5.5.2 Issue 2: Questionable Assumptions for Assigning Skin and Deep Dose for 

Unmonitored Workers Prior to 1960 by Means of Coworker Data 

As previously stated in the TBD: 

...From startup until July 1960, PGDP issued dosimeters to a limited number of 

individuals.... This population of monitored individuals represents those with the 

highest exposure potential. [Emphasis added.] 

Section 6.5 of the TBD provides guidance for assigning skin and deep dose to unmonitored 

workers prior to 1960 by defining the following three categories of unmonitored workers: 

Worker Group 1: A zero dose was recorded, but the worker was not monitored (majority 

of workers from 1953 to July 1960). 

Worker Group 2: A zero dose was recorded from the dosimeter system for any response 

less than the MDL. 

Worker Group 3: There was no recorded dose, because workers were not monitored or 

the dosimetry record is not available. 

Section 6.5.1 of the TBD provides the following “guidance” for assigning dose to unmonitored 

workers based on the “worker categories” described above. 

• 	 For Group 2, the missed dose should be based on n(MDL)/2 using data defined in 

Table 6-1 of the TBD. 

• 	 “If it is definitely established that the employee was not a radiation worker, then the 

unmonitored deep dose for the period can be assigned as the on-site ambient dose.” 

• 	 “Otherwise, an individual in Groups 1 or 3 should be treated as a radiation worker. The 

unmonitored dose can then be approached in two ways.” 

First, the same assignment of missed dose as for Group 2, from the last 

column of Table 6-1, can be considered. However, for the period 1953 

through July 1960, with the frequent (weekly) dosimeter exchange and 

relatively large MDL, the resulting implied annual missed dose of 1 rem is 

probably unrealistically large for many unmonitored persons in both 

Groups 1 and 3. Figure 6-2 shows the distribution of individual annual 

deep dose equivalent for monitored workers for the years 1953 to 1974.... 

Few of these individuals received as much as 1 rem in any given year. 

A second, alternative approach for Group 1 or 3 is to base the 

unmonitored dose estimate on exposure data compiled in (PACE/Utah 

2000) for monitored PGDP workers. The first four columns in Table 6-5 

(taken from Table 7.4 of the PACE report) show the number of monitored 
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workers, their average recorded deep dose, and the maximum individual 

deep dose for each year from 1953 through 1988. [Note: This table has 

been reproduced above as Table 5.5-2}. 

For the shallow dose, the TBD provides Table 6-6. For Groups 1 and 3, 

the TBD recommends the use of the GM and GSD for assigning dose to 

“unmonitored workers” for the period 1953 to 1960. 

For the large number of workers who were not monitored before 1960, guidance provided in the 

TBD for assigning dose is at best incomplete/confusing and subject to speculative 

decisionmaking. However, of greater concern are the underlying assumptions and interpretation 

of monitoring data, which provide the quantitative basis for assigning dose to unmonitored 

workers. 

It is SC&A’s opinion that the monitoring data cited in Table 6-5 of the TBD (and reproduced 

above in Table 5.5-2) have been misrepresented/misinterpreted, and the use of these data has the 

potential for significantly underestimating worker doses. At the center of this concern is 

NIOSH’s claim that prior to 1960, the population of “. . . monitored individuals represents those 

with the highest exposure potential.” 

It is SC&A’s contention that these monitoring data reflect a badging practice that not only 

included all worker categories (regardless of their potential for exposure), but furthermore 

diluted the average dose within a given worker category by rotating badge assignments. The 

following circumstantial evidence supports our contentions: 

• 	 Table 5.5.-1 above purports to provide average external cumulative penetrating dose and 

number of workers assigned to each department during 1953–1988. No fewer than 32 

different departments are identified with average cumulative doses ranging from 

3,814 mrem to a low of 110 mrem. 

In light of the claim that only workers with the highest potential for exposures were 

monitored, it is inconceivable that such selective monitoring would allow estimates for 

exposures back to 1953 in behalf of all 32 departments. For example, Table 5.5-2 

identifies a total of only 223 workers that were monitored in 1953. The ability to monitor 

such a broad spectrum of workers could only be achieved by cohort badging, in which a 

select few individuals for any given worker category are monitored. 

• 	 According to Table 5.5-2 between 1953 and 1960, the number of monitored workers 

increased from 223 to 526. After this time, all workers who entered a controlled 

radiation area were monitored. Table 5.5-2 identifies that for 1961, a total of 1,690 

workers were monitored. Thus, as part of this transition, an additional 1,164 workers 

were newly monitored workers. 

If, in fact, these newly monitored workers represented workers with low potential for 

exposures, the average worker’s dose for 1961 would have been expected to drop 

precipitously between the years 1960 and 1961. Inspection of Table 5.5-2, however, 
 

 

 

 

 

 

 

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shows that the average value of 201 mrem for 1960 dropped only 24 mrem to 177 mrem 

in 1961. Furthermore, this drop more than likely had nothing to do with the “expanded” 

monitoring practice that now included low-exposed workers, but reflected a general 

downward trend. Thus, in 1962, the average again dropped by 27 mrem, and in 1964 

dropped by 71 mrem. 

• 	 A 1964 Paducah Plant Health Physics Program report (Report No. KY-204, Rev. 3) 

authored by R.A. Winkel provides the following summary statistics, which are 

reproduced herein as Table 5.5-3 below. 

Table 5.5-3. Summary of Radiation Exposure from External Sources 

Paducah Gaseous Diffusion Plant (1953–1963) 

(Source: Winkel 1964) 

Whole-Body Number of Monitored Persons in Dose Range 

Dose Range Year 

(rad or rem) 1953 1954 1955 1956 1957 1958 1959 1960 1961 1962 1963 

0–1 220 

1–2 

273 389 

11 25 

418 

51 

624 

36 

631 

25 

554 

15 

506 

17 

1,609 

75 

1,418 

57 

1,261 

40 

2–3 3 10 8 5 2 4 5 4 8 

3–4 4 2 1 

4–5 1 

> 5 

Total Workers 220 284 417 484 670 662 571 527 1,689 1,479 1,309 

Data for the transition years of 1960 and 1961 clearly contradict the assumption/claim 

that prior to 1960, the monitored population represents workers with the highest potential 

for exposure. If that statement represented fact, then the addition of 1,162 newly 

monitored workers in 1961 would have been exclusively added to the 0–1 dose category. 

Instead, the data reveal that for 1960 and 1961, the number of workers with annual doses 

between 1 and 2 rem increased nearly 5-fold, from 17 to 75. 

In brief, these data further suggest that the post-1960 total rad worker population did not 

appear to differ from the pre-1960 monitored worker population, as claimed by NIOSH. 

It further suggests that the 57 additional workers with doses between 1 and 2 rem in 1961 

were not part of the monitored pre-1961 monitored worker group. Under current 

protocol prescribed by NIOSH, these 57 workers would likely be assigned the GM dose 

values for any years prior to 1961. 

• 	 The discrepancy between the claim of monitoring workers with the highest potential for 

exposure and the failure of dosimetry data to support this claim may very well be 

explained by yet another curious approach to monitoring workers. From a very limited 

body of available data, SC&A suspects that for the pre-1960 period, when badges were 

read on a weekly basis, workers may have been monitored on a “rotating basis,” as 

suggested by Exhibits #2, #3, #4, and #5. 
 

 

 

 

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Exhibits #2, #3, #4, and #5 represent excerpts from summary Inter-Company 

Correspondence for the Carbide and Carbon Chemical Company that operated the 

PGDP in 1953. These exhibits show the following: 

Exhibit #2: For the month of December 1952, 32 film dosimeters were used to 

monitor PGDP workers. 

Exhibit #3: For the second week of January 1953, 47 badges were assigned weekly. 

Exhibit #4: Between February 19 and March 18, 1953, weekly film dosimeter 

assignments increased from 56 to 77 badges. 

Exhibit #5: Over a 5-week period (i.e., from April 16 to May 20, 1953) “. . . the 

number of [weekly] film badges in use was increased from 77 to 96. 

These exhibits imply the following: 

(1) The total number of workers monitored on a weekly basis was small, but increased over 

time. 

(2) Locations selected from monitoring workers showed some variations. 

(3) The number of monitored workers was likely weighted by the total number of workers at 

each location but nevertheless represented only a fraction of the total number of workers 

at each location. 

(4) It appears that for any given location, worker monitoring was done on a rotational basis. 

In summary, data in Table 5.5-2 appear to be misleading. Table 5.5-2 identifies a total of 223 

workers as having been monitored in 1953. Implied in that statistic is that the same 223 

individuals were monitored every week, and that their GM dose represented an annual dose of 

140 mrem. The source for the average dose values in Table 5.5-2 is Table 7-4 of PACE/Utah 

(2000), where it is made clear that the average dose is that of the recorded doses. Those authors 

note that, “The large numbers of zero values in the database would reduce the average values 

below what they realistically may have been.” Similarly, the maximum dose is the maximum 

recorded value. The decision to use these values to represent the median and 95th percentile of a 

geometric distribution seems quite arbitrary. (Note that the GSD is a dimensionless quantity and 

should not have been labeled “rem” in Table 6-5 of Turner (2005).) 

Exhibits #2 through #5, however, suggest that in any given week, only a fraction of the total 

annual number of 223 workers were monitored on a rotating basis. For example, Person A may 

have been among the 223 monitored persons in 1953. However, it is possible that this worker 

may have been monitored for only 10 weekly cycles over the course of 52 weeks. Thus, records 

for this individual may show up to 10 positive weekly exposures, along with 42 or more zeros for 

cycles during which the worker was not monitored. 

If, as SC&A suspects, such a monitoring practice took place at PGDP, then all doses identified in 

Table 2 have been misrepresented, and are low and not claimant favorable. 
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5.5.3 Issue 3: Issues related to ORAUT-OTIB-0031 

ORAUT-OTIB-0031, External Coworker Dosimetry Data for the Paducah Gaseous Diffusion 

Plant (Merwin 2005), was issued for use on August 15, 2005, or 5 months after the issuance of 

ORAUT-TKBS-0019-6. The purpose of OTIB-0031 is to provide guidance for assigning 

external doses from penetrating and non-penetrating radiation in behalf of PGDP workers with 

no or limited monitoring data based on site coworker data. 

A comparison of ORAUT-OTIB-0031 with ORAUT-TKBS-0019-6 shows significant changes 

regarding the use of coworker data. These changes include (1) the use of “prorated” coworker 

doses that account for partial years of employment and/or incomplete monitoring records, and 

(2) the addition of missed doses for null dosimeter readings based on the formula n(LOD/2). 

As a result of these changes, coworker data that may be assigned to unmonitored PGDP workers 

under OTIB-0031 increased significantly. Table 5.5-4 compares the 50th percentile (or GM if the 

doses are from a lognormal distribution) deep dose coworker data cited in TKBS-0019-6 with 

revised coworker data as given in OTIB-0031 for the years 1953 through 1960. Inspection of the 

data shows that the revised values are several-fold higher than those cited in TKBS-0019-6. It 

appears that issues raised by SC&A pertaining to the coworker deep dose model in TKBS-0019­

6 have been resolved in OTIB-0031. Since both the 50th and 95th percentile doses are tabulated, 

the GSD for the corresponding lognormal distribution can easily be calculated. 

While similar adjustments were made for the non-penetrating dose in OTIB-0031, one 

outstanding issue that was not addressed in the TIB is the issue of the 80 mg/cm2

 attenuation of 

lower-energy betas. 

Table 5.5-4. Comparison of Median Annual PGDP External Coworker Deep Doses: 

TKBS-0019-6 versus OTIB-0031 

TKBS-0091061 OTIB-00312 

Year Pen. Dose (rem) Pen. Dose (rem) 

1953 0.1398 1.128
 

1954 0.2835 1.183
 

1955 0.2419 1.067
 

1956 0.3586 1.073
 

1957 0.2517 1.072
 

1958 0.1853 1.040
 

1959 0.2015 1.083
 

1960 0.2011 0.672 

1

 ORAUT-TKBS-0019-6, Table 6-5. 

2

 ORAUT-OTIB-0031, Table 2. 
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5.5.4 Issue 4: Issues Related to the Assessment of Neutron Exposures at PGDP 

The principal source of neutron exposure at PGDP involves the α, n reaction with fluorine 

compounds (UF4, UF6). This reaction would have the highest potential for exposure to workers 

in areas with storage cylinders that contained either depleted UF6 (tails) or enriched UF6. 

Other sources of neutrons involve uranium and transuranic elements that may undergo 

spontaneous fission and neutron-induced fission of fissile elements. 

Section 6.3.2.2 of ORAUT-TKBS-0019-6, however, states that “. . . the use of the commercial 

Harshaw thermoluminescent neutron dosimeters (TLND) to assess neutron dose routinely (along 

with deep and shallow dose) began [only] in 1998.” Thus, before 1998, workers were in effect 

unmonitored for neutron exposure. To account for missed neutron doses prior to 1998, Section 

6.5.3 of the TBD provides the following guidance: 

A neutron component should be added to the annual dose of individuals who 

worked in the cylinder yard before 1998. However, careful consideration should 

be given to work history. In general, only workers who were near cylinders for 

extended periods have the potential for neutron exposure. Estimates should be 

based on the neutron-to-photon ratio of 1 to 5 for dose equivalent, as determined 

from the survey conducted at PGDP (Meiners 1999). The neutron dose 

equivalent should then be multiplied by the ICRP (1990) factor of 2. [Emphasis 

added.] 

SC&A reviewed the referenced 1999 study by Meiners, Paducah UF6 Cylinder Painting Project 

Neutron Dose Study. In this study, TLD measurements were taken that assessed neutron-togamma dose ratios for radiation fields defined by depleted UF6 stored in cooled cylinders. Those 

measurements were compared to tissue-equivalent proportional counters. The study yielded 

neutron-to-photon ratios that ranged from 1:24 to 1:72, with an approximate average ratio of 1:5. 

The study also acknowledged the fact that the neutron-to-photon ratio is affected by the 

moderating/reflecting material in proximity to the cylinders, and the amount of UF6 contained in 

the cylinder(s), as well as the degree of uranium enrichment (an increase in U enrichment 

increases the alpha activity and, therefore, the α, n reaction). 

Lastly, SC&A reviewed guidance contained in ORAUT-OTIB-0024, Rev. 00, Estimation of 

Neutron Dose Rates from Alpha-Neutron Reactions in Uranium and Thorium Compounds, for 

consistency. Relevant data in OTIB-0024 are contained in Table 5-2, which identifies the 

neutron dose rate of 6.62E-10 rem/hr-gram of natural uranium without alpha-emitting progeny at 

a distance of 1 foot. For a large storage cylinder containing 100 lbs, a dose rate of 0.03 rem/hr at 

1 foot is calculated. This value is consistent with empirical measurements cited by Meiners 

(1999) and with guidance contained in ORAUT-TKBS-00190-6. 

Because workers were not monitored for neutrons prior to 1998, potential exposures to select 

workers have to be based on a neutron-to-photon dose ratio. The recommended neutron-tophoton ratio of 1 to 5 for estimating dose equivalent that is further multiplied by 2 to account for 
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the ICRP 60 neutron quality factor is regarded by SC&A as scientifically defensible for 

radiological environments defined by depleted UF6 in storage cylinders. 

However, a potential difficulty may arise for deriving unmonitored neutron doses for workers 

prior to 1961 who were unmonitored for photons as well. 
 

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6.0 OVERALL ADEQUACY OF THE SITE PROFILE AS A BASIS FOR 

DOSE RECONSTRUCTION 

The SC&A procedures call for both a “vertical” assessment of a site profile for purposes of 

evaluating specific issues of adequacy and completeness, as well as a “horizontal” assessment 

pertaining to how the profile satisfies its intended purpose and scope. This section addresses the 

latter objective in a summary manner by evaluation of (1) how, and to what extent, the site 

profile satisfies the five objectives defined by the Advisory Board for ascertaining adequacy; 

(2) the usability of the site profile for its intended purpose, i.e., to provide a generalized technical 

resource for the dose reconstructor when individual dose records are unavailable; and (3) generic 

technical or policy issues that transcend any single site profile that need to be addressed by the 

Advisory Board and NIOSH. 

6.1 SATISFYING THE FIVE OBJECTIVES 

The SC&A review procedures, as approved by the Advisory Board, require that each site 

profile be evaluated against five measures of adequacy; (1) completeness of data sources, 

(2) technical accuracy, (3) adequacy of data, (4) site profile consistency, and 

(5) regulatory compliance. The SC&A review found that the NIOSH site profile for 

PGDP presents an adequate accounting of the primary internal and external issues related 

to predominant radionuclides, such as uranium, plutonium, and some fission products. 

The PGDP site profile falls short in fully characterizing a number of key underlying 

issues that are fundamental to guiding dose reconstruction. In some cases, these issues 

may impact other site profiles. Many of the issues involve lack of sufficient conservatism 

in key assumptions or estimation approaches, or incomplete site data or analyses of these 

data. Section 6.0 summarizes the key issues. A detailed evaluation of these issues is 

provided elsewhere in the report. 

Objective 1: Completeness of Data Source 

The breadth of data sources used as a basis for the PGDP site profile is evident in the several 

hundred reports available for PGDP in the Site Profile Research Database. Over 100 reports 

were cited in the site profile references, while others served to provide confirmatory information, 

or were only recently retrieved. The NIOSH/ORAU team consulted health physics personnel 

with long histories at PGDP who have extensive knowledge of key dosimetry historical 

processes and personnel monitoring data. There was a meeting with the PGDP with the 

Construction Council and PACE on February 10 and February 11, 2005, respectively, in order to 

identify worker concerns and discuss the TBDs. This interaction has helped to provide valuable 

insight into site operations and processes. In addition, the issuance of supporting TIBs reflect the 

ongoing effort by NIOSH to continually improve guidance provided to dose reconstructors. 

The current version of the TBD is silent on the number of workers with zero dose, which makes 

it difficult to characterize the external dose distribution for the purpose of building a coworker 

model. Although the TBD has addressed major operations that occurred at PGDP, there are gaps 

with respect to the smelting operations, Work for Others program, onsite burning of 

contaminated material, and work supporting the weapons program. PGDP was integral in the 
 

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destruction of diffusion barriers removed as a part of the Cascade Upgrade Program for not only 

itself, but also for K-25 and the Portsmouth Gaseous Diffusion Plant. This introduces some 

uncertainty into the type of radionuclides handled and the relative concentrations of these 

radionuclides. 

Information in the TBD regarding the radionuclide composition of recycled uranium, as provided 

in the PACE report, is incomplete, which places constraints on the ability to reconstruct both the 

external and internal exposures. In addition, the list of transuranic radionuclides handled at the 

site appears to be incomplete. Table 5-5 of the TBD limits TRU to 237Np and 239Pu. However, 

as was discussed in the review of Table 5-4, the TRU in the Hanford reactor tails and other 

sources include 238Pu, 240Pu, and 241Am. These nuclides should be included with 237Np and 239Pu 

in Table 5-5, to be consistent with Table 5-4 (which lists all but 238Pu), SAIC (1999, p. 14), and 

Hill and Strom (1993, Table 16.2, Note “a”). Finally, a full disclosure of incidents at the site is 

not provided, nor are directions provided to dose reconstructors where incident data can be 

found. 

NIOSH did not make use of key information sources. The data in PACE/Utah 2000, which lists 

activity fractions for TRU and fission products associated with specific processes, was not 

adequately considered. A draft report on recycled uranium mass balance (BJC 2001), which is 

cited as a source of information by Turpin (2006), was also not adequately considered. In terms 

of environmental data, the USEC Characterization Study was not considered, although it 

provided an extensive characterization of the site prior to leasing the facilities to United States 

Enrichment Corporation. 

Objective 2: Technical Accuracy 

Several issues have been identified with regard to the accuracy of the dose reconstruction 

protocols recommended in the site profile. The principal technical accuracy issues include 

underestimating the enrichment levels of the uranium to which workers were exposed, and not 

addressing the complete list of TRU and fission products that were present in recycled uranium. 

In addition, for those TRU and fission products that are addressed, it appears that the quantities 

may have been underestimated. 

The TBD presents inadequate, and sometimes inaccurate, guidance to dose reconstructors on 

estimating the intakes of uranium, including the assumed specific activities of other 

radionuclides not analyzed by the bioassays. There is no guidance as to how the data on air 

concentrations can be used in dose reconstruction. Although we agree with the primary reliance 

on urinalysis, supplemented by in-vivo measurements where such records are available, 

measured air concentrations could be used to supplement the bioassay results when the bioassay 

records are missing or fragmentary. They could be especially useful in assigning relative 

specific activities of isotopes other than uranium. There is no discussion of the uncertainties in 

the determination of uranium in urine samples for fluorimetry and KPA. The uncertainty of the 

fluorimetry methodology can be very high—it is strongly dependent on the quenching of the 

sample. For other radionuclides, there is no description of the technique used for their 

measurements and the accompanying uncertainties. 
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The practice of offsite collection of samples that takes place 24- to 48-hrs after leaving the plant 

would lead to a lowering of the calculated intakes. The dose reconstructor is not instructed to be 

alert to any cases for which urine samples were collected after the worker was off work for any 

period of time. Appropriate adjustments to the calculated intake should be made to compensate 

for the lowered concentration in the urine following an absence from work. 

It is not clear if the TBD intends to give dose reconstructors the option of subtracting 0.43 μg 

from the 24-hr urine sample analyzed by techniques other than fluorimetry or KPA. Such 

guidance would be neither scientifically correct nor claimant favorable. First, the value cited by 

SAIC (1999) (the TBD refers to this document as BJC) applies to residents of the Oak Ridge, 

Tennessee, area, not of PGDP. Second, it is an upper limit (95th percentile). Subtracting such a 

high-end value from the urinalysis results would result in a bias against the claimant. The TBD 

should state unambiguously that the results should not be corrected for any assumed 

contributions of uranium in the environment. 

Table 5-10 presents the MDAs for the measurement of various radionuclides by whole-body or 

chest counting. The sources cited, however, do not present the values listed, with the sole 

exception of 237Np for the 1968–1991 period. Some of the values are inconsistent. The TBD 

gives no instruction to the dose reconstructor on estimating uncertainties in the in-vivo 

measurements, except to state, “For results near or at the reporting levels, dose reconstructions 

should apply the prescribed standard deviation of 0.3 times the MDA or reporting level...” 

While we do not disagree with that recommendation, clearer guidance is needed for higher 

levels. The uncertainties for in-vivo bioassay include the uncertainty in the counting statistics, 

including the background that is subtracted from the count rate; the uncertainty of translating a 

count rate from a given worker into a body burden, given the differences in morphology among 

different individuals; and the variation in the naturally occurring radionuclides in the human 

body (primarily 40K) that contribute to the background. These MDA values should be verified 

and documented to enable verification by an independent reviewer. 

Table 5-4 lists default specific activities to be used when only total uranium results are available. 

The report does not explain when this table should be used instead of Table 5-1, which gives 

percentages of uranium, neptunium, and plutonium activities at specific facilities, or Table 5-2, 

which presents default specific activities of uranium isotopes, as discussed above. This is a 

source of possible confusion for dose reconstructors. Not only do Tables 5-2 and 5-4 employ 

different units—Bq/μg and nCi/g, respectively—the unit conversions for the uranium isotope are 

imprecise. The specific activities of the uranium isotopes in Table 5-4, when converted to 

Bq/μg, are consistently about 3% higher than the default activities listed in Table 5-2. Although 

this is not a significant difference in terms of dose reconstruction, it is another example of the 

lack of accuracy and scientific rigor in the report. The table appears to be an amalgam of data 

from various sources—the report does not indicate how the values were calculated. 

Information about the occurrence of TRU nuclides at different facilities is presented by 

PACE/Utah (2000, Appendix D), which lists various radiological data. Air sampling, radiation 

survey, and urinalysis data indicates the presence of one or more TRU nuclides in buildings or 

facilities C-310, C-315, C-331, C-333, C-335, C-337, C-340, C-360, C-400, C-409, C-410, 

C-420, C-710, C-720, C-746, and C-746B. In addition, Hill and Strom (1993, Table 16.2) list 
 

 

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TRU nuclides at Buildings C-746A and “C-7460” (perhaps a typo for C-746Q). Finally, SAIC 

(1999, Table 1.7) lists TRU activities at C-333A, C-337A, C-404, C-411, and C-746Q. Since the 

dose assessments should always give the claimant the benefit of the doubt, the list of 

radionuclides in Table 5-5 should be all inclusive; therefore, the complete list of TRU nuclides 

should be included for each of these facilities. This requires expanding the radionuclide lists for 

C-310, C-315, C-337, C-333A, C-337A, C-340, and C-404. 

The particle sizes of at least some of the aerosols appear to be substantially smaller than the 

5 micron AMAD assumed as the default particle size throughout the TBD. The most claimantfavorable lung clearance types were not always recommended for the purpose of reconstructing 

internal doses. Recommended default assumptions regarding bioassay frequencies are not 

always claimant favorable, and no guidance is provided in the TBD to address the possible effect 

of the day of urine sample collection on the ability to reconstruct radionuclide intakes based on 

bioassay data. 

Tables presented with external monitoring summary data appear misleading. Table 5.5-2 

identifies a total of 223 workers as having been monitored in 1953. Implied in that statistic is 

that the same 223 individuals were monitored every week, and that their GM dose represented an 

annual dose of 140 mrem. The source for the average dose values in Table 5.5-2 is Table 7-4 of 

PACE/Utah (2000), where it is made clear that the average dose is that of the recorded doses. 

Those authors note that, “The large numbers of zero values in the database would reduce the 

average values below what they realistically may have been.” Similarly, the maximum dose is 

the maximum recorded value. The decision to use these values to represent the median and 95th 

percentile of a geometric distribution seems quite arbitrary. Information suggests that in any 

given week, only a fraction of the total annual number of workers were monitored on a rotating 

basis. 

The principal source of neutron exposure at PGDP involves the α, n reaction with fluorine 

compounds (UF4, UF6). This reaction would have the highest potential for exposure to workers 

in areas with storage cylinders that contained either depleted UF6 (tails) or enriched UF6. Other 

sources of neutrons involve uranium and transuranic elements that may undergo spontaneous 

fission and neutron-induced fission of fissile elements. Because workers were not monitored for 

neutrons prior to 1998, potential exposures to select workers have to be based on a neutron-tophoton dose ratio. The recommended neutron-to-photon ratio of 1 to 5 for estimating the dose 

equivalent that is further multiplied by 2 to account for the ICRP 60 neutron quality factor is 

regarded by SC&A as scientifically defensible for radiological environments defined by depleted 

UF6 in storage cylinders. 

The methods recommended for reconstructing shallow doses may result in an underestimate of 

doses, because no consideration was given to the methods used to calibrate the dosimeters. The 

coworker model for reconstructing doses to workers in the early years (i.e., pre-1960s) may be 

unreliable because of possible significant differences in work practices in the early years as 

compared to the later years where more abundant external dosimetry data are available. 

After discussion with NIOSH personnel, it was their decision to limit occupational medical 

exposure to those chest exams described above, and to conclude that all other exposures are part 
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of worker background. SC&A believes such an interpretation is not claimant favorable to those 

most at risk. Our concern is that specified “high-risk” workers, those most likely exposed to 

radiation and beryllium, would be at risk of having an incomplete dose assessment if not all 

radiation associated with medical screening for job-related activities were included. Since, all 

radiation provides some risk and, arguably, is cumulative, workers warrant consideration of all 

forms of work-related x-ray exposure to be claimant favorable. SC&A believes NIOSH should 

review its interpretation of included medical exposure, and should reasonably adopt a broader 

interpretation of occupational medical dose, as provided in the most recent version of Kathren 

and Shockley (2005). This is particularly important for skin cancer that is not one of the SECdefined cancers. The methods used to reconstruct occupational medical doses appear to 

underestimate the doses and do not characterize the full extent of the uncertainty in the doses. 

The TBD does little to reasonably document the variety of medical occupational exposures, and 

the lack of documentation on the type of equipment and the maintenance records do little to 

assure that a conservative and claimant-favorable estimation of dose is possible. This 

circumstance would suggest the need to reconsider a worst-case approach to establishing dose. 

The Occupational Medical Dose TBD in Section 3.2 states that PFG units, although generally 

available up to the late 1950s at most DOE sites, were not used at the PGDP. The undocumented 

absence of PFG units at PGDP clearly has significant dose implications to workers who may 

have been given much higher doses from PFG units. SC&A believes it is not claimant favorable 

to instruct dose assessors to use kerma (dose) values of 200 mrem and 100 mrem for chest 

radiography prior to 1975. To be fully claimant favorable, it would be appropriate to instruct 

dose assessors to use an annual dose of 3.0 rem per year for chest radiographs, in accordance 

with guidelines set forth (Kathren and Shockley 2005) until 1959, when you can assume all 

PFGs were no longer used at DOE sites. 

The TBD purports to describe “potential exposures from ambient sources while working outside 

the process buildings,” but the ambient monitoring data in the document are for measurements at 

the site boundaries and beyond. There are no corroborating data provided to demonstrate that 

these measurements are consistent with the levels that workers might experience while working 

outside the process buildings. Radionuclide-specific concentrations in air, soil, and water were 

not discussed in the TBD. Exposures from burning of contaminated materials were not 

considered in the assessment of ambient internal and external dose. There is little 

characterization of the releases in Tables 4-2 and 4-3. There is no discussion of what release 

points were considered, or the physical form of the uranium and 99Tc release data in Tables 4-2 

and 4-3, other than to indicate that they were obtained from DOE, Bechtel-Jacobs, and United 

States Enrichment Corporation reports. 

Section 4.3, External Dose, does not consider the radionuclides in the depleted uranium cylinders 

providing the gamma radiation for the increasing external exposure rates near the depleted 

uranium cylinder storage yards. Early survey results of about 0.02 mrem/hr (probably measured 

as 20 μR/hr) would correspond to 40 mrem in a 2000-hr period, or about 175 mrem/yr for 

continuous exposure. For comparison, the average of the 1993–2001 reported background 

measurements in Table 4-4 correspond to a 100 mrem/yr rate for continuous exposure. The 

uncertainty for environmental dose deals less with estimating the uncertainties in measured or 
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calculated values than with outlining default values and assumptions to be used when data are 

unavailable. 

Objective 3: Adequacy of Data 

There appears to be a paucity of external dosimetry data prior to the early 1960s, and the 

reconstruction of external doses during those early years requires the application of a coworker 

model. Similarly, it appears that reliable neutron monitoring did not begin until 1998. As a 

result, reconstruction of neutron exposures from alpha/n reactions and spontaneous fissioning 

must rely on a coworker model. Data gaps also exist in the environmental dose TBD, because 

the data are limited to measurements made at the site boundary and do not include external 

exposure measurements performed onsite at worker locations. 

It was the policy of PGDP when film badge and urine sampling showed a “lack of value” that 

these services be discontinued (Emlet 1957). No explanation for the term “lack of value” is 

provided. This determination to discontinue these services introduced a potential opportunity for 

unmonitored exposures. During site expert interviews, there was indication that not all 

individuals participated in in-vivo monitoring, and that a cohort-monitoring program was 

utilized. A quantitative neutron-monitoring program was not implemented until 1998. Prior to 

this date, there was no documented neutron dose. These conditions should be considered during 

the dose reconstructions. 

The Occupational Medical Dose TBD is also deficient in that little documentation exists to 

validate x-ray protocols and equipment maintenance, and upkeep of records prior to 1995, after 

which the State of Kentucky commenced surveys of the unit. The TBD uses information derived 

from the TIB (Kathren 2003) to estimate dose impacts. 

Objective 4: Consistency among Site Profiles 

An extensive comparison was performed by SC&A to compare and contrast the methodologies 

used in the PGDP site profile and other site profiles reviewed to date. These comparisons focus 

on the methodologies and assumptions associated with dose assessments and the derivation of 

values used to obtain a probability of causations (POC) for individual claimants. A detailed 

analysis is provided in Attachment 4 to this report. 

The default values assigned for determining medical exposure are relatively consistent among 

site profiles. The site profiles do not always apply the same revision of this ORAUT-OTIB0006, as is the case with the ORNL site profile. This should be corrected in subsequent revisions 

of the TBDs. A more consistent approach to determining when dose from PFG is to be assigned 

is needed in cases where there is an absence of site-specific information. Other deviations from 

the standard assumptions are based on site-specific information. 

The default assumptions used for the calculation of environmental occupational dose are the 

same or similar to other site profiles. The PGDP TBD assumed an intake calculation based on a 

chronic intake over a year, with a breathing rate of 2,400 m3

 of air in a year. External ambient 

exposure was based on gamma exposure rates on mrem by location. An occupancy of 
 

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2,000 hours per year was used. This is consistent with other site profiles, such as LANL, ORNL, 

and others. Review of site profiles to date indicates that the NIOSH/ORAU team has not come 

to a consensus on what components should be considered in the environmental dose. In the case 

of the PGDP Occupational Environmental Dose TBD, no consideration was given to dose from 

contaminated soil resuspension considered in the RFP site profile. 

There is an inconsistency within the TBD related to the type of dosimeters used between 1953 

and 1980. In one location, the TBD indicates a “two-element” dosimeter was used, while 

Table 6-1 of the TBD only identifies four-element film dosimeters in use between 1953 and 

1980. This “technical” discrepancy has significant implications when viewed in context with the 

statement that the resultant beta dose is likely to be claimant favorable, “...because of the 

significant over-response of the unshielded portion of the film to any lower energy photons that 

have been present.” 

At PGDP, whole-body counting was performed using a mobile counter provided by the Y-12 

Plant that was sometimes referred to as the Martin Marietta Energy Systems (MMES) Counter. 

Table 5-10 on page 18 of the PGDP Occupational Internal Dose TBD (Berger 2004) lists general 

information about the detection capabilities of this counting system for various periods, and 

provides MDAs for various operational periods at PGDP. In the Fernald Occupational Internal 

Dose TBD (Rich 2004, pg. 34), Table 5-22 provides a similar table of MDA for the same mobile 

counter, which shows that the MDA for non-uranium radionuclides was 0.1 pCi/sample 

throughout most of the years. Uranium in-vivo MDAs for the lung at Fernald are provided in 

Table 5-26 on page 36. Whereas the MDA at PGDP was 4 mg for U-238 from 1968 to 1980 

(Berger 2004, Table 5-10, pg. 18), the MDA for U-238 from 1989 to 2001 at Fernald was 7.4 mg 

(Rich 2004, Table 5-26, pg. 36). The lung counting detection levels at Y-12 using two 9” x 2.5” 

NaI detectors varied from 13.5 mg in 1959 down to 4.5 mg when the HP germanium detectors 

were installed (Rich and Chew 2006, Table 5-12, pg. 32). 

Exposure geometry is not dealt with in the PGDP Occupational External Dose TBD. At 

ORNL, the Occupational External Dose TBD (Burns and Mohrbacher 2004, pg. 23) 

points out that 100% Anterior-Posterior (AP) exposure geometry has been assumed. The 

Occupational External Dose TBDs for Y-12 (Murray 2003), SRS (Scalsky 2005), and 

Hanford (Scalsky 2003) base their default exposure geometry on the compensability or 

non-compensability of the claim. The MCW (Westbrook 2005) and RFP (Furman and 

Lopez 2004) Occupational External Dose TBD bases default exposure geometries on job 

titles. The INEEL (Rohrig 2004) Occupational External Dose TBD defaults to 100% AP 

exposure. 

Objective 5: Regulatory Compliance 

SC&A reviewed the site profile with respect to Objective 5, which requires SC&A to evaluate 

the degree to which the site profile complies with stated policy and directives contained in 

42 CFR Part 82. In addition, SC&A evaluated the TBDs for adherence to general quality 

assurance policies and procedures utilized for the performance of dose reconstructions. NIOSH 

has complied with the hierarchy of data required under 42 CFR Part 82 and its implementation 

guides. 
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The TBDs’ use of personnel monitoring data and environmental monitoring data to determine 

dose is consistent with the requirements outlined in 42 CFR Part 82, as follows: 

• 	 Where in-vivo and in-vitro analyses are available, this information is provided for use in 

determination of internal dose. 

• 	 Where routine beta/gamma and neutron dosimeters are available and adequate, this 

information is provided for use in determination of external exposure. 

• 	 Where environmental measurements are available, these data are used as the basis for 

environmental dose. 

The particle size in each case is specified as 5 μm AMAD. Such a particle-size assignment is not 

supported by the data, nor is it claimant favorable. This decision to ignore site-specific data is 

contrary to both the intent and the letter of 42 CFR 82. 

6.2 USABILITY OF SITE PROFILE FOR INTENDED PURPOSES 

SC&A has identified seven criteria that reflect the intent of the EEOICPA and the regulatory 

requirements of 42 CFR Part 82 for dose reconstruction. Because the purpose of a site profile is 

to support the dose reconstruction process, it is critical that the site profile assumptions, analytic 

approaches, and procedural directions be clear, accurate, complete, and auditable (i.e., 

sufficiently documented). SC&A used the following seven objectives to guide its review of the 

PGDP site profile TBDs to determine whether it meets these criteria: 

Objective 1 − Determine the degree to which procedures support a process that is expeditious 

and timely for dose reconstruction 

Objective 2 − Determine whether procedures provide adequate guidance to be efficient in select 

instances where a more detailed approach to dose reconstruction would not affect the outcome 

Objective 3 − Assess the extent to which procedures account for all potential exposures, and 

ensure that resultant doses are complete and are based on adequate data 

Objective 4 − Assess procedures for providing a consistent approach to dose reconstruction, 

regardless of claimants’ exposures by time and employment locations 

Objective 5 − Evaluate procedures with regard to fairness and the extent to which the claimant is 

given the benefit of the doubt when there are unknowns and uncertainties concerning radiation 

exposures 

Objective 6 − Evaluate procedures for their approach to quantifying the uncertainty distribution 

of annual dose estimates that is consistent with and supports a Department of Labor POC 

estimate at the upper 99% confidence level 
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Objective 7 − Assess the scientific and technical quality of methods and guidance contained in 

procedures to ensure that they reflect the proper balance between current/consensus scientific 

methods and dose reconstruction efficiency 

Inconsistencies and Editorial Errors in the Site Profiles 

Several inconsistencies and editorial errors were noted in the TBD and are provided below. 

• 	 Americium-241 is misidentified as a decay product of 239Pu. The TBD (pg. 7) states 

“...241Am, a decay product of 239Pu, builds up as 239Pu decays.” Americium-241 is not a 

decay product of 239Pu; it is a decay product of 241Pu (EPA 1999, Table G.1). 

• 	 The air sampling flow rate characterization is confusing. Section 4.2.2 states that 

according to the September 1962 Air Sampling Procedure, the sampling flow rate was 

0.3 cfm (11 cm/s). The 11 cm/s value is the face velocity and not a flow rate equivalent 

to 0.3 cfm, as the sentence implies. 

• 	 The predominant wind direction in Section 4.2.2 is misstated. The wind rose in 

Figure 4-1 shows the frequency of winds blowing from the indicated directions. The text 

on pg. 7 of Section 4.2.2 incorrectly identifies the wind as blowing predominantly 

toward the southeasterly direction, rather than as predominantly from the southsouthwest direction. 

• 	 There appear to be some discrepancies between Tables 4-3 and 4-4 and their descriptions 

in the text. In Table 4-2, alpha concentrations are assigned for 1952–1957, 1990, and 

1993 (rather than 1952–1956, 1990, and 1994, as described in Section 4.2.4). In Table 4­

3, beta concentrations are assigned for 1954–1960, 1995, and 1997–2001 (rather than 

1952–1958, 1989, 1993, and 1995–2001, as described in Section 4.2.4). The assigned 

values are not consistent with the calculations described. 

• 	 The first paragraph of Section 5.2: “Source Term,” singles out 230Th and 234mPa as 

uranium progeny “of dosimetric interest.” This designation of 230Th is apparently an 

error; DOE (2000) cites 234Th and 234mPa, the short-lived progeny of 238U. 

These errors may cause confusion in the dose reconstruction process and should be modified in 

subsequent revisions of the TBD. 

6.3 UNRESOLVED POLICY OR GENERIC TECHNICAL ISSUES 

A number of issues were identified that are common in the PGDP and other site profiles 

reviewed to date and, in some cases, represent potential generic policy issues that transcend any 

individual site profile. These issues may involve the interpretation of existing standards (e.g., 

oro-nasal breathing, metal tritides, high-fired plutonium oxide), how certain critical worker 

populations should be profiled for historic radiation exposure (e.g., construction workers and 

early workers), and how exposure itself should be analyzed (e.g., treatment of incidents and 

statistical treatment of dose distributions). NIOSH indicated that it may develop separate TIBs 
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in order to address these more generic issues. The following represents those issues identified in 

the PGDP and previous site profile reviews that, in SC&A’s view, represent transcendent issues 

that need to be considered by NIOSH as unresolved policy or generic technical issues. 

(1) Direction on the applicability of generic TIBs to individual dose reconstructions for 

particular sites is absent. 

(2) Mobility of the work force between different areas of the site should be addressed. Site 

expert testimony that many workers moved from one plant to the next is a complicating 

factor. Establishment of an accurate worker history is crucial in such cases. This will be 

especially difficult for family-member claimants. 

(3) Statistical techniques used in the application of the data to individual workers should be 

further considered and substantiated. 

(4) Dose from impurities and/or daughter products in radioactive material received and 

processed at sites should be assessed as a contributory exposure source. 

(5) The significance of various exposure pathways and the assumptions made that influence 

dose contributions need to be considered (most notably) for solubility, oro-nasal 

breathing, and ingestion. 

(6) Analysis needs to be performed regarding how “frequent or routine incidents” should be 

addressed, given the possibility that such “spike” exposures often may be missed by 

routine monitoring as a function of how often and in what manner it was conducted. 

(7) Availability of monitoring records for “transient or outside workers,” e.g., subcontractors, 

construction workers, and visitors, who may have potential exposure while working on or 

visiting a facility should be ascertained. 

(8) Dose to decontamination and decommissioning (D&D) workers should be assessed. 

Many facilities have large-scale D&D operations, which extend back many years. 

Decontamination and decommissioning operations often require working in unknown 

situations, which may provide unique exposure situations. 

(9) Dose from non-traditional chemical forms of radionuclides, such as high-fired oxides and 

tritides, requires evaluation. 

(10) Dose reconstruction for occupational medical exposures remains incomplete. NIOSH 

needs to reconsider the definition to include all forms of medical radiation exposure to 

ensure its considerations are claimant favorable. 

Quality Assurance on records provided by the site to the NIOSH/ORAU team is necessary to 

ascertain whether complete information is being provided. 
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NCRP (National Council on Radiation Protection and Measurements ), 1989. Medical X-Ray, 

Electron Beam and Gamma-Ray Protection up to 50 MeV Equipment Design, Performance and 

Use), NCRP Report No. 102, Bethesda, Maryland. 

NIOSH (National Institute for Occupational Health and Safety), 2002, External Dose 

Reconstruction Implementation Guidelines, OCAS-IG-001, Revision 1, Cincinnati, Ohio. 

NRC (National Research Council), 1989, Film Badge Dosimetry in Atmospheric Nuclear Tests, 

National Academy Press, Washington, DC. 

PACE/Utah 2000, (Paper, Allied Industrial, Chemical and Energy (PACE) Workers International 

Union and University of Utah Division of Radiobiology Center for Advanced Medical 

Technologies, Center of Excellence in Nuclear Technology, Engineering and Research), 

Exposure Assessment Project at The Paducah Gaseous Diffusion Plant. Sponsored by U.S. 

Department of Energy, Office Environment, Safety and Health, 2000. 

Pacific Northwest National Laboratory, Environment, Safety, and Health Systems (PNNL). 

2004. “Hanford Internal Dosimetry Technical Basis Manual,” PNNL-MA-860. 

http://www.pnl.gov/eshs/pub/pnnl860.html 

Patterson, G.R., 1958, Evaluation and Control of Internal Exposure From Enriched Uranium at 

Y-12, in Symposium on Occupational Health Experience and Practices in the Uranium Industry, 

(HASL-58), U.S. Atomic Energy Commission, Health and Safety Laboratory, New York, NY, 

October 15-17, 1958. 

PHS (U.S. Public Health Service), 1970, Radiological Health Handbook, Rockville, Maryland. 
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Rich, B.L., 2004, Technical Basis Document for the Fernald Environmental Management Project 

(FEMP) – Occupational Internal Dose, ORAUT-TKBS-0017-5, Rev. 00, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee, May 28, 2004. 

Rich, B.L. and Chew, M.H., 2006, Technical Basis Document for the Y-12 Site – Occupational 

Internal Dose, ORAUT-TKBS-0014-5, Rev. 02, Oak Ridge Associated Universities, Oak Ridge, 

Tennessee, August 30, 2006. 

Rohrig, N., 2004, Technical Basis Document for the Idaho National Engineering and 

Environmental Laboratory (INEEL) – Occupational Medical Dose, ORAUT-TKBS-0007-3, 

Rev. 00, PC-1, Oak Ridge Associated Universities, Oak Ridge, Tennessee, May 28, 2004. 

SAIC (Science Applications International Corporation) 1999. “Internal Dosimetry Technical 

Basis Document for Bechtel-Jacobs Company LLC,” BJC/OR-258. 

SC&A 2004. SC&A Standard Operating Procedure for Performing Site Profile Reviews, Rev. 0, 

S. Cohen and Associates, McLean, Virginia. March 18, 2004. 

Scalsky, E.D. 2003. Technical Basis Document for the Hanford Site – Occupational Medical 

Dose, ORAUT-TKBS-0006-3, Rev. 00, Oak Ridge Associated Universities, Oak Ridge, 

Tennessee. September 16, 2003. 

Scalsky, E.D., 2004. Technical Basis Document for the Savannah River Site To Be Used for 

EEOICPA Dose Reconstructions, ORAUT-TKBS-0003, Revision 2, October 29, 2004, Oak 

Ridge Associated Universities, Oak Ridge, Tennessee. 

Scalsky, E.D., 2005, Savannah River Site, ORAUT-TKBS-0003, Revision 3, Oak Ridge 

Associated Universities, Oak Ridge, Tennessee, April 5, 2005. 

Scott, L.M., et al., 1969, Design and Development of a Mobile In vivo Radiation Monitoring 

Laboratory, American Industrial Hygiene Association Journal, March-April 1969. 

Smith, R.F. 1984. “Historical Impact of Reactor Tails on the Paducah Cascade,” KY/L-1239. 

Tomes, T., 1997, Technical Basis for Internal Dosimetry at the FEMP, SD-2008, Rev. 1, Fluor 

Fernald, Inc., Fernald, Ohio. 

Turner, James E. 2004, Technical Basis Document for the Paducah Gaseous Diffusion Plant – 

Occupational Medical Dose, ORAU-TKBS-0019-3, (Rev. 00), ORAU Team Dose 

Reconstruction Project for NIOSH, 11 February 2004. 

Turner, James E., 2005, Paducah Gaseous Diffusion Plant – Occupational External Dose, 

ORAUT-TKBS-0019-6, (Rev. 01), ORAU Team Dose Reconstruction Project for NIOSH, 29 

March, 2005. 

Turner, J. E., and Thomas B. R., 2006, Technical Basis Document for Paducah Gaseous 

Diffusion Plant – Occupational External Dosimetry, ORAUT-TKBS-0019-6, Rev. 02, Oak Ridge 

Associated Universities, Oak Ridge, Tennessee, August 30, 2006. 
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Turpin, Baynard, 2006, Technical Basis Document for the Paducah Gaseous Diffusion Plant – 

Site Description, ORAUT-TKBS-0019-2, (Rev. 01), ORAU Team Dose Reconstruction Project 

for NIOSH, 21 August 2006. 

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2%22 

UCC (Union Carbide Corporation) 1949, Health Physics Progress Report: January 1 through 

31, 1949. Oak Ridge Tennessee, February 14, 1949. 

“Urinalysis Program.” n/d. ORAUT ID: Paducah Gaseous Diffusion Plant\06-29-04 – Data 

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“Uranium in Urine Analysis.” n/d. ORAUT ID: Paducah Gaseous Diffusion Plant\04-28­

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USEC (United States Enrichment Corporation) 1994, The Uranium Enrichment Process, Rev. 4, 

United States Enrichment Corporation, October 1994. 

Warner, W.T., March 16, 1966, letter to C. L. Carl, Report of UF6 Gas Release on February 14, 

1966, Feed Materials Production Center, Fernald, Ohio. 

Westbrook, J.L. 2005. Basis for Development of an Exposure Matrix for the Mallinckrodt 

Chemical Company St. Louis Downtown Site and the St. Louis Airport Site, St Louis, Missouri, 

Period of Operation, 1942–1958, ORAUT-TKBS-0005, Rev. 01, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee. March 10, 1005. 

Wiley, S.W., 2002, Letter to Mr. James Newton, April 10, 2002. 

Winkel, R.A. 1964, Paducah Plant Health Physics Program Report (No. KY-204, Rev. 3). 

Wolf, B.S., 1945, X-ray Department at Y-12, as of 23 February 1945, Memorandum for the files, 

Army Service Forces, United States Engineer Office, Manhattan District, Oak Ridge, Tennessee, 

February 23, 1945. 
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ATTACHMENT 1: RICHARD MILLER LETTER TO HHS AND NIOSH 

Letter from Richard Miller to the Secretary of the Department of Health and Human Services 

and the Director of the National Institute for Occupational Safety and Health 
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ATTACHMENT 2: TECHNICAL DOCUMENTS CONSIDERED DURING 

THE REVIEW 

• 	 Technical Basis Document for Paducah Gaseous Diffusion Plant – Introduction, 

ORAUT-TKBS-0019-1, Rev. 00 PC-1 (Maisler 2006). 

• 	 Technical Basis Document for the Paducah Gaseous Diffusion Plant – Site Description, 

ORAUT-TKBS-0019-2, Rev. 01 (Turpin 2006). 

• 	 Technical Basis Document for the Paducah Gaseous Diffusion Plant – Occupational 

Medical Dose, ORAUT-TKBS-0019-3, Rev. 00 (Turner 2004). 

• 	 Technical Basis Document for the Paducah Gaseous Diffusion Plant – Occupational 

Environmental Dose, ORAUT-TKBS-0019-4, Rev. 00 (East 2004). 

• 	 Technical Basis Document for Paducah Gaseous Diffusion Plant – Occupational Internal 

Dose, ORAUT-TKBS-0019-5, Rev. 00 (Berger 2004). 

• 	 Technical Basis Document for Paducah Gaseous Diffusion Plant – Occupational 

External Dose, ORAUT-TKBS-0019-6, Rev. 01 (Turner 2005). 

These documents are supplemented by two technical information bulletins (TIBs), which provide 

additional guidance to the dose reconstructor: 

• 	 Ikenberry, Tracy A., 2005. Internal Dosimetry Coworker Data for Paducah Gaseous 

Diffusion Plant, ORAUT-OTIB-0037, Rev 00, ORAU Team Dose Reconstruction Project 

for NIOSH. September 20, 2005. 

• 	 Merwin, Steven E., 2005. External Coworker Dosimetry Data for the Paducah Gaseous 

Diffusion Plant, ORAUT-OTIB-0031, Rev 00 PC-1, ORAU Team Dose Reconstruction 

Project for NIOSH. August 15, 2005. 
 

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ATTACHMENT 3: SUMMARY OF EXPERT INTERVIEWS 

To be provided upon clearance. 
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ATTACHMENT 4: CONSISTENCY AMONG SITE PROFILES 

The default site profile assumptions and methodologies for Paducah Gaseous Diffusion Plant 

(PDGF) are summarized below and were compared to those of other site profiles reviewed to 

date or in current review. Site profiles completed to date by the SC&A team include Bethlehem 

Steel, Mallinckrodt Chemical Works (MCW), Iowa Army Ammunition Plant (IAAP), Hanford, 

the Savannah River Site (SRS), the Y-12 National Security Complex (Y-12 Plant), Idaho 

National Engineering and Environmental Laboratory (INEEL), the Nevada Test Site (NTS), and 

the Rocky Flats Plant (RFP). Additional site profiles in the process of review are Oak Ridge 

National Laboratory (ORNL), the Mound Plant (Mound), and Fernald (FEMP) and Los Alamos 

National Laboratory (LANL). ORNL had multiple missions which overlap with a number of 

other sites in the DOE complex such as weapons research, the heat source program, reactor 

research, assembly and disassembly operations, and tritium operations to name a few. 

To ascertain the differences in assumptions between what assumptions are used for the PGDP 

site profile versus other site profiles, the assumptions from each PGDP TBD must first be 

understood. The core assumptions for each TBD have been outlined below. 

Dose Reconstruction Assumptions for Occupational Medical Exposure 

The Paducah Occupational Medical Dose TBD (Turner 2004) has relatively good information on 

the site specific data related to x-ray equipment and techniques used at the Paducah Gaseous 

Diffusion Plant (PGDP) which is summarized in a table and discussed in the TBD. 

Table 3-2 lists the diagnostic medical X-ray equipment used at PGDP during 

specified periods. The initial General Electric (GE) machine was used from the 

opening of the plant in 1952 through February 1975. It was replaced by the 

Picker unit, which served from March 1975 through December 1995. 

The present equipment has been in operation since January 1996. Quality 

assurance (QA) has been verified regularly by the Food and Drug Administration 

and the State of Kentucky, as well as by inhouse surveys. Interviews with the staff 

provided much of the information in this TBD. The X-ray facility has been 

operated by the present technician since November 1974. (Turner 2004, pg. 5) 

Table 3-2 on page 6 provides information on types of filtration, collimators, Bucky grids and 

film development processors. Table 3-3, also on page 6 of the TBD, provides further 

information by time periods on the type of x-ray (PA and LAT), the kVp (V) for each type of 

film, and the current, which in all cases is 300 mA. 

From March 1975 to the present, the dose with either of the two more recent 

machines listed in Table 3-2 has been comparable for a given procedure. 

Therefore, organ dose equivalents are determined for two periods: 1952 to 

February 1975 and March 1975 to the present. (Turner 2004, pg. 6) 

Table 3-9 on page 12 of the TBD discussed the potential sources of uncertainty in organ dose 

equivalent assessment, discussed conservative default values (referring to Kathren et al. 2003), 
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and refers to Table B.3 in Report NCRP 102 (NCRP 1989) for a list of air kerma per mAs at 

different distances. ICRP 34 (ICRP 1982) is referred to for organ-absorbed doses per unit 

entrance kerma. 

In the last row of Table 3-9, the same conversion factors from entrance kerma to 

organ doses are used for all individuals, a distinction being made between male 

and female for some organs. In any case, the conversion factors are 

representative for an exposed individual (for the assumed kerma) to the extent 

that the anatomical features of the individual match those of the phantom on 

which the tables are based. The variation introduced in this step is not known. 

An indication can be gained through comparison with dose conversion factors in 

ICRP (1982) for the 5-year-old pediatric phantom under the same irradiation 

conditions. (Turner 2004, pg. 12) 

The Fernald Occupational Medical Dose TBD (Chew 2004) has more specific information for 

the dose reconsrtructor in Section 3.0 of the TBD. 

Only exposure to X-rays that were performed as a condition of employment is 

included; X-rays that resulted from injuries or medical testing are not included. 

Medical exposures are assigned or assumed based on the actual or calculated 

frequency of chest X-rays required for employment. 

There is no difference between the exposures experienced by unmonitored versus 

monitored workers since dosimeters normally worn by workers were not worn 

during diagnostic medical X-ray examinations. All exposures are external based 

on the assumption that internal radioisotopes were not administered for 

diagnostic purposes (with the exception that some radioisotopes might have 

been used at specific sites for stress testing as part of working conditions). 

Specific organ doses to be attributed for posterior-anterior (PA) chest X-rays 

should be calculated on the basis of the dose conversion factors found in 

International Commission on Radiological Protection (ICRP) Publication 34 

(ICRP 1982). 

Organ doses from lateral chest radiography should be estimated at 2.5 times 

greater than those from the corresponding PA doses, based primarily on the 

greater mAs exposure per radiograph and the somewhat smaller source-to-skin 

distance (SSD). If technique factors can be identified for Type I equipment, organ 

doses for Type I equipment may be calculated by multiplying the organ dose 

estimates for Type II equipment by 2.5. This approach is reasonable when 

compared to other U.S. Department of Energy (DOE) sites (e.g., Hanford) where 

more information on X-ray equipment from the early period is available. 

For organs not listed in ICRP (ICRP 1982) but specified in the Interactive 

RadioEpidemiological Program (IREP) code, doses should be determined by 

analogy with anatomical location. Using this logic, IREP code organs in the 

thoracic cavity that are not mentioned in ICRP (ibid) can be assigned 
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the same dose as the lungs; doses to organs in the head and neck can be assigned 

the same dose as the thyroid. Head and neck organ (i.e., eye and brain) dose 

estimates should be somewhat greater than doses actually incurred (and therefore 

are claimant-favorable) because of geometry considerations and, at least in the 

case of the brain, because of attenuation by the bony cranium. To ensure 

claimant-favorability in view of the variations in organ dose described in ICRP 

(ibid, p. 51), apply the dose values for females, which are slightly higher than 

those for males. 

For both PA and lateral views, a standard source-to-image distance (SID) of 

72 inches (183 cm) may be assumed unless specifically noted otherwise. 

If not specified, assume that all X-ray machines were single-phase and that there 

was no air gap between the patient and the film. 

If there is an indication that mobile X-ray units were used, this could result in 

higher doses. This would imply that this type of unit (which could have been a 

photofluorographic machine) might have delivered higher exposures. (Chew 

2004, pg. 5) 

The Fernald Occupational Medical Dose TBD spends a whole page describing the type of x-ray 

equipment and provides a summary Table 3-3 on page 12 of the types of x-ray equipment used at 

FEMP. Table 3-4 (Chew 2004, pg. 12) provides detailed information on a 1961 survey of the 

x-ray units. Later surveys in 1995 and 1997 provided much additional information and assisted 

in learning how to decrease dose to the workers. 

A 1995 survey by the DHHS noted a significant increase in the entrance skin 

exposure over the 1993 survey (i.e., from 15.3 mR to 28.0 mR). The FEMP 

response was that since the 1993 inspection (and following a risk/benefit 

analysis) FEMP had switched (for approximately 1 year) to the Kodak InSight 

Thoracic Imaging System. This system was designed to improve overall clinical 

performance related to chest examinations. It enabled FEMP to obtain 

significantly more usable information during a single session in comparison to 

the former “conventional” imaging system. The new system required 

changes in equipment settings (i.e., the technique) to accommodate and allow for 

the best diagnostic utilization of the technically superior film. The “increased” 

measured ESE of 28 mR using the new technology was still below the 

recommended Federal ESE guideline of 30 mR for routine chest 

radiography. 

A 1997 DHHS survey found that the measured ESE was 32.2 mR. The technique 

used was a manual mode at 110 kVp and 8 mAs. The survey noted that the film 

optical density (OD) was outside the defined range. DHHS recommended that 

FEMP contact the service representative for the X-ray system and the Kodak film 

representative to discuss methods to lower patient ESE, and change exposure 

techniques from a manual to a phototimed process. FEMP made the 

recommended changes to improve the OD, and in a return inspection DHHS 
 

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measured the ESE to be 19.2 mR. The technique used was 4 mAs at 126 kVp. For 

the assessment of organ doses during this period, the higher ESE of 32.2mR was 

used. (Chew 2004, pg. 15) 

The Y-12 Occupational Medical Dose TBD (Murray 2006a) points out that the type of x-ray 

equipment is not known for the period from 1948 to 1968. A description of what is known about 

the x-ray equipment at Y-12 is listed in Table 3-2 on page 8, as well as a more detailed 

discussion of the types of x-ray equipment on pages 8 and 9. 

According to information provided by the Y-12 Medical X-Ray Department, preemployment chest X-rays were always taken with a conventional medical 

diagnostic X-ray machine (Wiley 2002). They found no evidence of PFG chest 

X-rays in the employee medical X-ray folders and all chest X-rays in these folders 

were 14”x17” films. However, in reviewing the medical X-ray folders of workers 

from the 1940s, approximately 1400 4”x10” chest X-ray films were found in the 

medical X-ray folders of workers who were employed at Y-12 from 1943 to 1947. 

Originally these X-rays were thought to be copies of conventional X-rays that 

were taken elsewhere and sent to Y-12 when the person was hired there. In fact, 

these were PFG chest X-rays. 

In February 1945, a General Electric (GE) stereoscopic photoroentgen unit is 

listed as an equipment item in the Y-12 Medical Department (Wolf 1945). 

Reexaminations and other chest films were done on conventional 14”x17’ films. 

On October 12, 1945, the Tennessee Eastman Corporation (TEC) Medical 

Director sent a telegram to the GE X-Ray Corporation, requesting them to set up 

a Photoroentgen Unit 4x10 (Leggo 1945). In June 1946, TEC placed an order 

with the Oak Ridge Hospital for 6000 Eastman-Single Coated 4”x10” X-ray films 

for the period from August 1, 1946 to July 31, 1947 (Graham 1946). Thus, it is 

clear that pre-employment chest X-rays were taken with a PFG unit from 1943 to 

1947, as evidenced by the 4”x10” films found in the medical records and the 

purchasing records mentioned above. All chest X-rays done since then are 

conventional 14”x17” X-rays. 

No record has been located to determine what type of diagnostic X-ray machine 

was in use at Y-12 from 1948 until the GE-type machine mentioned above that 

was used in the early 1960s. This date may not be correct either. In a meeting 

with the X-ray technologist who provided the information for that report, the 

technologist said that more recent documentation indicated that the GE-type 

machine was installed in 1969, not in the early 1960s (Beck 2003). Thus, it is not 

possible at this time to state with certainty what X-ray machine was used from 

1948 to 1968. From 1969 until January 1982, the X-ray machine in use was 

similar to a GE Model DXD-350. This machine was replaced with a GE Model 

DSX-650II in February 1982 (Wiley 2002). 

A description of the X-ray equipment used at Y-12 is included in Table 3-2. The 

specific technique factors for these machines are shown in Table 3-3. Since no 
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technique factors were identified by Y-12 for Types I and II equipment, organ 

doses, based on assumed technique factors, were developed on the basis of X-ray 

techniques contemporary with the time period (1943–1968), with due 

consideration given to claimant favorability. (Murray 2006a, pg. 8) 

The Y-12 Occupational Medical Dose TBD (Murray 2006a) devotes a section to providing the 

information needed by the dose reconstructor, but some of the needed information is not 

available requiring the development of default values. 

No actual X-ray output measurements are available. The X-ray technique factors 

provided may not be reliable, especially for the Type II equipment. Thus, default 

values for entrance kerma will be used in the calculation of organ dose 

conversion factors (DCFs) for use in dose reconstruction. Default values have 

been developed for the three of the most commonly used occupational medical 

diagnostic X-ray procedures: PA, lateral; and PFG chest films (ORAU 2003).. 

The default values are considered to be maxima developed from reviews of patient 

doses reported in the literature, machine characteristics, and knowledge of X-ray 

procedures used during the time periods indicated. Sufficient conservatism was 

included in the determination of the default values to ensure with near certainty 

(>99% confidence) that the actual exposures from the specified procedures would 

not exceed the default values, thus ensuring claimant favorability. 

In determining the default factors in Table 3-4, it was assumed that minimum 

filtration was used, along with low kilovolt peak (kVp) techniques, slow film 

speeds with standard development procedures, and no additional collimation or 

use of cones. The default entrance kerma values for the three procedures 

are given in Table 3-4 (Murray 2006a, pg. 9) 

A source-to-image distance (SID) of 72 inches (in) (183 centimeter [cm]) was 

standard for the time for the PA chest, and 42 in (106 cm) for the PFG chests. 

The X-ray machines used at Y-12 were most likely single-phase, and typically no 

air gap was used between the patient and the film. Before 1982, it is assumed that 

the X-ray equipment was operated at 80 kVp, had at least 1.5 millimeters (mm) 

aluminum (Al) total filtration (see Table 3.1 of National Council on Radiation 

Protection and Measurements [NCRP] [1989]), and that the half value layer 

(HVL) was approximately 2.5 mm Al equivalent [eq.] (see Table B.2 of NCRP 

[1989]). These were typical machine parameters for chest X-rays performed in 

this time period. After 1982, the X-ray equipment was operated at 110 kVp and 

had at least 2.5 mm Al total filtration. The HVL was approximately 3.5 mm Al eq. 

After 1982, the machine parameters were the same but the exposures were phototimed (Wiley 2002). The default values for entrance kerma were also used for the 

PA chest X-rays after 1982 because the exposure time would not be known for a 

photo-timed (automatic) exposure. (Murray 2006a, pp. 9–10) 
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The default values for Occupational Medical Dose are the same as those used in other site 

profiles as noted on the following pages of the Occupational Medical Dose TBD (Turner 2004). 

• 	 Exam included posterior-anterior (PA) and limited use of lateral (LAT) views (pg. 5). 

• 	 Dose conversion factors were taken from International Commission on Radiological 

Protection (ICRP) Publication 34 (ICRP 1982) in Sections 3.4.3 and 3.4.4 to evaluate the 

dose equivalent in various organs (pg. 7). The skin is the only organ listed in Table 3-7 

that does not involve ICRP (1982) dose conversion factors (pg. 10). For 1952–February 

1975, the dose conversion factors of Kathren et al. (2003) for pre-1970 are used (pg. 20). 

• 	 The air kerma and resultant organ doses are proportional to the time integrated current 

(mAs) (pg. 6). 

• 	 Default values for skin-entrance kerma have been developed for use in such instances by 

Kathren et al. (2003) (pg. 8). 

• 	 As in (Kathren et al. 2003), a beam quality HVL of 2.5 mm Al was assumed, and 

substitute projections were used for some organs to approximate the lack of good 

collimation (pg. 9). 

• 	 Total filtration is equivalent to 2.5 mm Al. The air kerma value Ko = 0.19 

cGy/(100 mAs) for single-phase generators at a source-to-image distance (SID) ro = 183 

cm (= 72 in) (pg. 7). 

• 	 Air gaps between the patient and the film (not addressed in the TBD). 

• 	 During the period from March 1975 to the present, when both PA and LAT views were 

made, the posterior skin (Table 3-7) received a dose equivalent of 140 mrem (with 

backscatter) from the PA view plus radiation entailed from the LAT view (without 

backscatter) (pg. 10). 

• 	 In the absence of other information, initial organ dose equivalents HLS to be used for 

lumbar spine dose reconstruction are estimated to be those from all five views, roughly 

approximated as 2.5 times the sum of the AP and LAT dose equivalents. Values of HLS 

are given in Table B-1 (ICRP 1982) (pg. 19). 

• 	 “For estimating dose equivalents with the Interactive RadioEpidemiological Program 

(IREP) for organs not included in ICRP (1982), these organs are classified in three 

anatomical regions, as listed in the first column of Table 3-7. In the second column, a 

single organ from ICRP (1982) is selected from Table 3-6 as representative of the dose to 

all organs in that region. Column three lists other body organs according to the region 

in which they are located” (pp. 9–10). 
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• 	 Rough estimate of uncertainty for x-ray procedures is 50% (ICRP 1982) (Table 3-9, 

pg. 12). 

If one looks at the consistency between the Paducah, Fernald and Y-12 Occupational Medical 

Dose TBD the following summary helps to put this into perspective. 

At Paducah, the frequency of x-ray examinations is covered in Table 3-1 (Turner 2004, pg. 5). 

For non-smokers, and smokers up to 40 years of age, it is every 5 years. For non-smokers over 

40 years of age, it’s every three years. For asbestos workers, it’s every two years. At FEMP 

(Chew 2004, Table 3-2, page 11), it’s annually during the time period, 1952 to 1981 and from 

1982–2002, for employees over 45 it is annually, but employees under 45 are offered a x-ray 

every two years. For Y-12 (Murray 2006a, Table 3-1, pg. 7), the frequency changed during four 

time periods, 1943–1988, 1988–1993, 1993–1998, March 1988 to present. Although, during the 

early period, 1943–1988 x-rays were taken annually for pre-placement, annual and termination 

examination. Form 1988 to 1993, chest x-rays changed based on age from annually to every 10 

years. Starting in 1993, Y-12 annual chest-rays were only given to asbestos and beryllium 

workers and annual exams were not taken, in general, unless it was for a preplacement or 

termination examination. 

The PGDP (Turner, pg. 10) and Fernald (Chew 2004, Table 3-6) site profiles refer to ICRP 34 

(ICRP 1982) for organ dose conversion factors. The Fernald Occupational Medical TBD 

provides an additional Table 3.7 (Chew 2004, pg. 10) that provides dose conversion factors that 

are not provided in ICRP 34. The Y-12 Occupational Medical Dose TBD (Murray 2006a, pg. 

19) provide two tables, Table 3C-2 (1943–1947) and Table 3C-3 (1948–1968) that provide dose 

conversion factors for the various organs. 

The Fernald Occupational Medical Dose TBD (Chew 2004, Table 3-1, pg. 8) provides a 

consolidated summary of the relationship of beam intensity and various technical factors. A 

simplified table like this does not appear in the Paducah Occupational Medical Dose TBD, 

although a table of this type is helpful to the dose reconstructor. 

Whereas the Y-12 Occupational Medical Dose TBD (Murray 2006a, pg. 17) defines the IREP 

radiation type as photons, from 30–250 keV, the Paducah and Fernald Occupational Medical 

Dose TBD do not spell this out clearly. 

The Paducah Occupational Medical Dose TBD (Turner 2004, pg. 12) provides Table 3-9 that 

discussed potential sources of uncertainty in organ dose equivalent assessments that is most 

helpful to the dose reconstructor. On page 11 (Turner 2004), it mentions that the uncertainty is a 

positive ± 30%. The Fernald Occupational Internal Dose TBD (Chew 2004, pg. 8) in Table 3-1 

lists uncertainty as ± 30%, assuming all errors are positive. The Y-12 Occupational Medical 

Dose TBD (Murray 2006a, pg. 12) assumes an uncertainty of ±30 % at the 99% confidence level 

may be assumed. For further conservatism, it may be appropriate to assume that errors are all 

positive, and only the + 30% should be used. The Y-12 Occupational Medical Dose TBD also 

has a whole Section 3.5 on page 11 that discusses uncertainty. 
 

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Dose Reconstruction Assumptions for Occupational Environmental Exposure 

The Paducah Occupational Environmental Dose TBD (East 2004) provides an overview and 

describes the potential exposures to Paducah Gaseous Diffusion Plant (PGDP) workers from 

ambient sources while working outside the process buildings. Section 4.2 of the TBD also 

discusses internal dose from the breathing of airborne concentrations of radionuclides released 

while on the PGDP site. Section 4.3 describes potential external dose from sources of radiation 

outside process buildings. Section 4.4 discusses possible sources of uncertainty associated with 

both of these sources of environmental dose. Some of the major assumptions made in the TBD 

are included below. 

Using the maximum airborne radionuclide concentration in a year (either from 

actual data or an estimated value), annual intakes for the radionuclides of 

concern were derived by using an assumed annual respiration rate of 2,400 m3

/yr. 

Most years have only a potential for contributing only about 1 mrem CEDE or 

less. Intakes from inhalation are potentially significant only for operations before 

1963. (East 2004, pg. 10) 

Unmonitored workers in the early years did not have significant inventories of 

depleted uranium to contribute to external dose. Later, unmonitored workers 

would not spend their entire work year at the depleted cylinder storage yards and, 

therefore, would not reach the maximum dose recorded by fence line monitoring. 

No other significant sources of external exposure are associated with the PGDP 

operations. An assumed deep dose equivalent rate of 200 mrem/yr for all years 

would be reasonable, and deficiencies in earlier measurement techniques thereby 

become immaterial. (East 2004, pg. 12) 

All external environmental dose data were adjusted to reflect a 2,000-hr work 

year. The data were originally reported in site environmental reports as 

representative of an employee who worked at the site 24 hours a day, 365 days a 

year. Using an employee permanently on the site, however, is an unrealistic 

assumption that would clearly overstate onsite environmental exposures. (East 

2004, pg. 15) 

Uncertainty related to internal exposures presents similar concerns. The highest 

internal exposures occurred during the early years, and lessened over time with 

increased controls and better equipment. Assumption of the maximum uptake for 

all years reduces the need to include an uncertainty factor (other than default 

values) for intakes. (East 2004, pg. 15) 

In summary, external exposure rose over time as the depleted uranium inventory 

grew, and internal exposure decreased as releases were reduced, providing 

offsetting factors to ensure claimant favorability. Therefore, the external dose 

equivalent that can be applied though the history of PGDP is 200 mrem/yr. 

Reconstructions should estimate annual intakes using the information in 

Section 4.2.5. (East 2004, pg. 15) 
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This TBD relies on several data sources for estimating ambient airborne radionuclide 

concentrations and for annual airborne releases by radionuclide. The principal sources were the 

annual environmental reports for PGDP from 1958 and through 2001. Measurements from the 

air-monitoring locations have been provided in Tables 4-2 and 4-3 and applied using the method 

described in the TBD. Attachment 4A provides an expanded version of this data, including 

additional background monitoring points. 

The Fernald Occupational Environmental Dose TBD (Chu 2006) describes the type of 

environmental exposure at the Feed Materials Production Plant (FEMP) by stating the following: 

....the primary radiation exposure pathways are inhalation of airborne 

radionuclides and exposure to direct radiation from emission plumes, radioactive 

materials in the process plants, and radioactive substances deposited on the 

ground or surfaces. Dose reconstructors can determine the internal dose from 

inhalation of radioactive materials from radionuclide air concentrations, and the 

external dose from exposure to radioactive materials outside the body from 

historical data of external dosimetry monitoring. (Chu 2006, pg. 8) 

The TBD provides in Table 4-4 a list of annual airborne emissions due to thorium operations at 

Plants 8 and 9 and the Pilot Plant from 1951 through 1988. The TBD points out that thorium 

process emissions represent the best approximation possible given the lack of specific production 

information and assumes that, at best, this estimate is accurate within an order of magnitude. 

Because of the conservative assumptions used for scrubber and dust collector efficiencies and the 

intake of material to the collection equipment, the TBD cautions the dose reconstructor that 

actual thorium emissions should not exceed those listed in Table 4-4. 

Some of the assumptions made in the Fernald Occupational Environmental Dose TBD include 

the following. 

The average thoron activity concentration in outside air is comparable to the 

environmental background concentration of radon; therefore, it can be assumed 

that FEMP thoron environmental concentrations vary with fluctuations in radon 

concentrations (Tomes 1997) (Chu 2006, pg. 10) 

...for 1972 and subsequent years, an activity concentration equal to that of the 

radon background concentration is assigned to areas close to buildings where 

thorium was stored. (Chu 2006, pg. 10) 

This TBD analysis assumed that the radioactive aerosol concentration varies 

significantly in various sections of the Plant. The dose received by a worker is 

highly dependent on the amount of time spent in specific work areas. (Chu 2006, 

pg. 10) 

The intake calculation assumed that an individual would breathe in 2,400 m3

 of 

air in a year. The concentration and annual intake values for uranium and 
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thorium in Table A-1 are in milligrams per cubic meter and milligrams 

respectively. Table A-4 lists specific activities and isotopic content (or activity 

fraction of an isotope) for converting milligrams to Becquerel of individual 

isotopes. Radon (222Rn) and thoron (220Rn) concentration values are in pCi/L. 

(Chu 2006, pg. 20) 

For 222Rn, 1 WL= 100 pCi/L and the derivation of WLM is based on an individual 

who continuously breathed air at the FEMP for 2,000 hr for the year 

and an assumed environmental radon daughter product equilibrium factor of 

70%. [This ratio for ambient outside air is in accordance with widespread 

sampling conducted throughout the United States that is referenced in NCRP 

(1984)]. For 222Rn, 1 WL=7.47 pCi/L and the equilibrium factor have been 

documented to range between 0.02 and 0.1 at several locations (Tomes 1997). 

(Chu 2006, pg. 20) 

The radionuclide release rates used to estimate EA concentrations were the 

average of those release rates calculated for AMS 8 and 9, using ground release 

(χ/Q)s. The source of release was assumed to be at the center of the Production 

Area (Plant 4). EA concentrations and intakes were calculated using these input 

data. (Chu 2006, pg. 25) 

The Y-12 Occupational Environmental Dose TBD (Murray 2006b) point outs that there are two 

exposure pathways for environmental dose; (1) inhalation of uranium in ambient air due to 

operational releases, and (2) direct external radiation from radionuclides in soils and outdoor 

surfaces, as well as shine from buildings and operational units. The TBD states that there are 

two potential sources of external exposures received by workers at the Y-12 facility: 

(1) exposures from the deposition of radionuclides released as a consequence of facility 

operations, and (2) exposures received from radiation levels emanating from buildings and 

storage areas. 

Due to the complexity of the terrain surrounding Y-12 and the release 

mechanisms from the production facilities compounded by the limited dispersion 

distances, traditional dispersion and transport models were considered 

unsuitable. An empirical approach based on the limited ambient air monitoring 

at Y-12 was used to estimate uranium air concentrations. This approach used the 

annual release estimates independently reconstructed by previous studies to 

generate annual air concentrations for four locations within Y-12. (Murray 

2006b, pg. 9) 

The ORDR Project estimated releases from 1944 to 1995. Thus, this source 

cannot be used to estimate air concentrations from 1996 to 2002. The release 

estimates in Table 4.2.4-1 show a definite downward trend for the years 

preceding 1995 (Figure 4.3-2). Thus, it is conservative to assume that the air 

concentrations reported from 1996 to 2002 are equal to the concentrations 
 

 

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reported for 1995. Air concentrations and intakes for the four on site locations 

are presented in Tables D-1 to D-4 in Attachment D. The uranium air 

concentrations are based on the 50th and 95th percentile value of the 

lognormally distributed Chi/Q. (Murray 2006b, pg. 29) 

Estimates of intake assume 2,000 h y-1 exposure. An inhalation rate of 1.2 m3 h-1 

generates an annual intake off 2,400 m3 y-1. (Murray 2006b, pg. 30) 

The on site concentrations of 234/

235U and 238U were estimated based on an 

empirical approach. The air concentrations and intakes were estimated using the 

50th and 95th percentile values for the empirical dispersion coefficient (Chi/Q) 

for each station. Air concentrations (becquerel/cubic meter [Bq m-3]) for all 

years from 1948 to 2002 were estimated based on the quantities of uranium 

released. Intakes (Bq y-1) were estimated from the air concentrations based on 

an exposure assumption of 2000 h y-1 and an inhalation rate of 1.2 m3 h-1.d 

(Murray 2006b, pg. 42) 

A comparison of the assumptions made with respect to environmental dose from the Paducah 

Occupational Environmental Dose TBD (East 2004) are compared below with those in the 

Fernald and Y-12 Occupational Environmental Dose TBDs. 

• 	 In the Paducah Occupational Environmental Dose TBD (East 2004) it is pointed out that 

measurements from the air-monitoring locations listed in Tables 4-2 and 4-3 were 

collected from documents and applied using a method described in the TBD. 

Attachment 4A provides an expanded version of this data, including additional 

background monitoring points. Table 4-4 provides PGDP external gamma exposure rate 

of mrem/2,000 hr by year and location. 

• 	 In the Fernald Occupational Environmental Dose TBD (Chu 2006) Attachment D shows 

background concentration measurements. The intake calculation assumed that an 

individual would breathe in 2,400 m3

 of air in a year. The concentration and annual 

intake values for uranium and thorium in Table A-1 are in milligrams per cubic meter and 

milligrams respectively. 

• 	 The Y-12 Occupational Environmental Dose TBD (Murray 2006b) provides the 

following information about measurements. 

No measurements of external exposures were reported for the twelve on 

site stations at Y-12, which began operations in 1983. With the exception 

of the limited data from environmental monitoring reports, there are two 

major characterizations of external exposures that have been performed 

for Y-12. A series of aerial radiological surveys was performed in 1973– 

1974, 1980, 1989 and 1992 for the ORR that included Y-12. These 

surveys consisted of a reservation wide, high level survey and low-level 

facility specific surveys. The second major characterization was 
 

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performed from 1985 to 1987 and involved an outdoor radiological and 

chemical scoping survey of the 800+ acres occupied by the Y-12 facility. 

External environmental dose equivalent rates outside buildings on Y-12 site from 1948–2002 are 

provided (Murray 2006b, Table 4.4.4-1, pg. 33). 

• 	 The Paducah Occupational Environmental Dose TBD (East 2004) indicated that PGDP 

had established by 1958 a network of permanent stations on and off the site to collect 

continuous ambient air samples. Sampling data are available in annual environmental 

reports published since 1958 for four perimeter locations (inside the fence) and varying 

numbers of offsite locations. The Paducah TBD assists the dose reconstructor by 

providing Table 4A-1 (pg. 21) that provides outdoor beta concentrations (Bq/m3 ) as 12 

monitoring stations for each year from 1952 to 2001. Likewise for outdoor alpha 

concentration, the TBE provides a Table 4A-2 (pg. 22) that provides outdoor beta 

concentrations (Bq/m3 ) as 12 monitoring stations for each year from 1952 to 1996. 

• 	 Similarly in the Fernald TBD (Chu 2006, pg. 9) it is reported that Table 4-1 lists annual 

airborne uranium emissions due to FEMP operations throughout the operating history 

(1951 through 1988) from each production plant and processing facility. The Fernald 

TBD has an Attachment A, which provides an airborne radionuclide concentration and 

intake summary that lists the radionuclide concentration and intake summary by year for 

1951 to 1988 for uranium, thorium, and Rn-222. (Bq conversion factors are provided in 

Table A-4). The TBD also provides an Attachment C, which lists the FEMP radionuclide 

concentration and intake results by exposure area and year. 

• 	 The Y-12 TBD (Murray 2006b, pp. 44–50) discusses occupational environmental doses 

in Attachment D. Attachment D provides (in separate sections) calculated 234/235U and 

238U air concentrations and intakes for Stations 2, 4 8 and 12; site-wide 234/235U and 238U 

air concentrations and intakes based on average air concentrations for Stations 2, 4, 8 and 

12; maximum 234/235U and 238U air concentrations and intakes; and external dose rates 

outside buildings on Y-12 site. 

Dose from the resuspension of contaminated soil has not been considered in the current version 

of the Paducah Occupational Environmental Dose TBD (East 2004). It is hoped that NIOSH is 

considering further investigation into this route of exposure. Soil resuspension dose was 

considered in relation to environmental exposure at both the Rocky Flats Plant (McDowellBoyer and Little 2004) and the SRS (Scalsky 2005). It appears that further investigation into 

environmental source terms is needed. The NIOSH/ORAU team is encouraged to continue the 

screening of source terms and the inclusion of these results in the TBD. 

Dose Reconstruction Assumptions for Occupational Internal Exposure 

The Paducah Occupational Internal Dose TBD, ORAUT-TKBS-0019-5 (Berger 2004) describes 

the default assumptions for occupational internal dose at PGDP. The assumptions were derived 

from historical records relating to the in vivo, the in vitro, and the air monitoring programs. 
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Assumptions from the Paducah Occupational Internal Dose TBD (Berger 2004) are noted 

below; page numbers are from the TBD: 

A review of in-house procedures used to assess the concentration of uranium in 

urine indicates that a variety of quality control steps were an integral part of the 

process. For example, duplicates were consistently run, and comparison of 

results to known quantities was a critical step. Therefore, the in vitro results from 

in-house processing, typically reported in units of micrograms of uranium per 

liter, can be considered generally reliable. However, interpretation of those 

results can be difficult, primarily because of uncertainty about enrichment, 

solubility, and the contribution of environmental uranium, but also because 

samples were collected at work and during the middle of the workweek, meaning 

that cross-contamination and the inability to separate soluble from insoluble 

intake fractions contribute to the uncertainty. 

Nonetheless, dose reconstructors can prepare reliable estimates of dose from the 

dates of employment, the employment locations, and the urine bioassay results. 

(In vivo results, because they were acquired primarily in response to an incident, 

are less reliable for assessing routine intakes.) Assumptions such as absorption 

types and the presence or absence of TRU elements can be derived from the 

historical records. (Berger 2004, pg. 6) 

For dose assessment purposes, a nominal distribution of radionuclides must be 

assumed because not all analytical methods were capable of detecting many of 

the radionuclides in the PGDP source term. If only total uranium results are 

available for a particular measurement result, Table 5-4 provides a default 

isotopic distribution. (Berger 2004, pg. 7) 

As of the date of this report, there is little information to support a reduction in 

the types or forms of radionuclides listed in this table on an annual basis, and few 

effective decommissioning efforts that could have resulted in downgrading these 

radiologically restricted areas. Therefore, it is assumed that all of the 

radionuclides shown were present in the buildings beginning in 1953. Before 

1953, reactor tails were not used as feed to the cascade, and the TRU materials 

do not apply. (Berger 2004, pg. 12, footnote a, Table 5-5) 

Unless site-specific information is available, the particle size is assumed to be 5 

μm AMAD, as recommended in ICRP (1994, paragraph 5). (Berger 2004, pg. 12, 

footnote c, Table 5-5) 

In some cases, a detection level for a particular radionuclide or analysis method 

is not available. In that case, dose reconstructors should use the nominal 

detection levels in Table 5-8. In addition, if it is not clear from the monitoring 

records how/where a particular claimant’s sample was analyzed, it should be 

assumed that they were analyzed in-house (i.e., at the PGDP) and the MDC from 

that measurement type used to assess missed dose. Finally, if a record contains a 
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notation of “less than X micrograms/L” or “< x pCi/sample”, that value should 

reflect the MDC for that sample. (Berger 2004, pg. 15) 

Urine samples were typically collected in the workplace at PGDP, predominantly 

on Wednesdays. Therefore, contamination of samples from the worker’s hands or 

clothing cannot be ruled out as a contributor to any given result. If a second 

analysis was performed and if that result was negative, it is reasonable to assume 

the first result was a false positive due to sample contamination or laboratory 

error. (Berger 2004, pg. 15) 

 A footnote to the above paragraph provided the following additional information: “At some 

point, the sample collection day changed to Monday. However, the date of that procedural 

change is unclear.” 

MDA = minimum detectable activity or amount. The MDAs shown for uranium, 

while given in units of mass, are presumed to have been based on measurement of 

Th-234 activities along with an assumed isotopic ratio. It is presumed that the 

results for enriched uranium are based on measurement of the measurement of 

U-235 activities and an assumption of enrichment. However, these presumptions 

cannot be confirmed thus cautious use of these MDA values is recommended. 

(Berger 2004, pg. 18, footnote b, Table 5-10. In-vivo measurement types and 

detection levels for various periods.) 

For in vivo measurements, contamination could have occurred as external to the 

body or, in the case of chest counting, as external to the lung. If a follow-up in 

vivo count (the same day or within a few days) showed a dramatic decrease in 

activity or no detectable activity, then external contamination should be assumed. 

(Berger 2004, pg. 18) 

NIOSH has developed a technical information bulletin to assist in dose reconstruction for 

unmonitored workers by using coworker data. OTIB-0037, Internal Dosimetry Coworker 

Data for Paducah Gaseous Diffusion Plant (Ikenberry 2005) includes a number of 

assumptions on intake modeling as noted below. 

The IMBA Expert OCAS-Edition computer program requires urine results to be in 

units of activity per day. The total uranium results are in units of μg/L; therefore 

the results were multiplied by 1.4 in order to normalize them to the Reference 

Man excretion rate of 1400 mL per day. (Ikenberry 2005, pg. 6) 

Bioassay results were converted from mass to activity before fitting assuming 

0.0389 Bq/μg, characteristic of low-enrichment (2 percent) uranium. Lowenrichment uranium feed is the default value in ORAU [Berger 2004] when the 

specific location where a claimant worked is not available. (Ikenberry 2005, 

pg. 6) 
 

 

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The uncertainty for each result in the intake calculation was assumed to be 

normally distributed. All results were equally weighted by applying a uniform 

absolute error of 1, indicating to IMBA that all results are (assumed to be) 

equally precise. (Ikenberry 2005, pg. 6) 

A chronic exposure pattern was assumed for PGDP workers; this pattern also 

approximates a series of acute intakes with unknown intake dates. Intakes were 

assumed to be by inhalation using a default breathing rate of 1.2 m3/hr and a 

5-μm activity median aerodynamic diameter (AMAD) particle size distribution. 

(Ikenberry 2005, pg. 6) 

The database reported all results as “uranium.” The bioassay data were assumed 

to represent excretion of 234U. (Ikenberry 2005, pg. 6) 

All uranium isotopes considered have long half-lives in relation to the assumed 

intake period so radioactive decay is not a consideration. Also, all uranium 

isotopes are biokinetically identical so there is no effect on the fitting of the data 

for intake determination. 234U was the isotope selected because it would result in 

the highest internal dose; the International Commission on Radiological 

Protection (ICRP) Publication 68 dose coefficients (also referred to as dose 

conversion factors) for 234U are 7% to 31% larger than those for 235U, 236U, and

238U (ICRP 1995). (Ikenberry 2005, pg. 6) 

Because of the isotopic compositions of the source terms, the assumption of 234U 

will yield claimant-favorable doses. PGDP received uranium and began 

enrichment operations during June 1952 and first withdrew enriched uranium 

during November 1952. The November 1952 period is consistent with uranium 

urinalysis data; however, the first intake period was conservatively assumed to 

begin on June 1, 1952. (Ikenberry 2005, pg. 6) 

The lognormal distribution is selected in the Interactive RadioEpidemiological 

Program (IREP), with the calculated dose entered as Parameter 1 and the 

associated GSD as Parameter 2. The GSD is associated with the intake, so it is 

applied to all annual doses determined from the intake period. (Ikenberry 2005, 

pg. 9) 

The Fernald Occupational Internal Dose TBD, ORAUT-TKBS-0017-5 (Rich 2004) describes the 

default assumptions for occupational internal dose at the Fernald Environmental Management 

Project (FEMP). The assumptions were derived from historical records relating to the in-vivo, 

the in-vitro, and the air monitoring programs. 
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Assumptions from the FEMP Occupational Internal Dose TBD (Rich 2004) are noted 

below; the page numbers are from the TBD: 

On February 14, 1966 an accidental release of approximately 1200 kg of uranium 

occurred during conversion operations that resulted in elevated personnel 

exposures (Warner 1966). The enrichment is unknown and 2% should be 

assumed for any claimant identified as a subject of this incident. In addition, 

more than 70% of the thorium at FEMP was handled and processed from 1964 to 

1979 in the Pilot Plant. (Rich 2004, pg. 8) 

In the absence of specific enrichment information, and considering the above 

available data related to processing experience of uranium enrichments at 

FEMP, the default assumption for time periods after 1964 is 2% enrichment for 

bioassay data in milligram quantities of uranium. Prior to 1964 natural 

uranium should be assumed. (pg. 10) 

Table 5-3 lists the primary assumptions for FEMP uranium enrichments and the 

isotopes associated with these enrichments. The mass percentages, relative 

activities in pCi/μg, and the total pCi/μg values are based upon IMBA NIOSH 

default values. (pg. 10) 

Uranium -1% enriched (EU assumption used in early in vivo calculations) 

(pg. 10, Table 5-3). 

Uranium–2% enriched (Recommended dose evaluation default for this TBD) 

(pg. 10, Table 5-3). 

U-236 is less than 1% activity in DU, Ntl, 1% EU, and 2% EU. (pg. 10, Table 5-3) 

In the absence of specific enrichment information, and considering the above 

available data related to processing experience of uranium enrichments at 

FEMP, the default assumption for time periods after 1964 is 2% enrichment for 

bioassay data in milligram quantities of uranium. Prior to 1964 natural 

uranium should be assumed. (Rich 2004, pg. 10) 

Before 1989 no TRU analyses for radiological safety were performed on a routine 

basis for either airborne or urine activity, and exposure controls remained based 

on chemical toxicity under the assumption that these controls would be sufficient 

for all the radiological issues (Bassett et al. 1989). Although the alpha activity 

from the TRU alpha emitters would have been collected and detected on the air 

samples, the reported results were all considered to be uranium and compared to 

the MAC. (Rich 2004, pg. 14) 

Before February 1989, no smears or air sampling filters were analyzed 

specifically for plutonium, neptunium, or thorium isotopes (Basset 1989), 

although these radionuclides would have been detected by gross alpha counting. 

In 1989, several sets of air and surface smear samples from Plants 4 and 8 were 
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analyzed for total uranium, 239Pu/240Pu, 238Pu, 237Np, 228Th, 230Th, and 232Th. 

Table 5-8 lists the results in derived air concentrations (DAC) from these 

analyses. DAC ratios are used not for dose reconstruction purposes, but only to 

indicate the measured relative levels of TRU contaminants in the plant in the 

1989 time period, which in turn lends credence to the default assumptions in 

Table 5-8 for accounting for the unmeasured TRU in the plants. (Rich 2004, 

pg. 15) 

The chemical forms of the RU contaminants are not known, although it is 

apparent from the chemical processes to which the materials were subjected 

during uranium processing, a variety of forms would be expected. Hence the dose 

reconstructor should assume the most claimant favorable solubility type for the 

target organ. (Rich 2004, pg. 17) 

For the purposes of dose reconstruction, effective equilibrium is a logical, 

claimant-favorable assumption because thorium was present from the earliest 

times and was stored at FEMP after the industry need for thorium products was 

past. However, for purposes of perspective Figure 5-1 presents thorium 

equilibrium growth curves. (Rich 2004, pg. 19) 

The ICRP has assigned oxides and hydroxides of thorium to inhalation type S. All 

other compounds of thorium are type M. The claimant-favorable assumption 

would be either type M or S, based on the organ of interest, because all of the 

compounds in Table 5-14 were handled and could have resulted in intakes. The 

default isotopes would be 228Th and 232Th in equilibrium, since the degree of 

equilibrium is impossible to estimate due to the variable times of separation of the 

isotopes from the feed stock (see Figure 5-1). Table 5-15 lists the estimated 

thorium emissions from FEMP plant facilities in comparison with uranium 

emissions. (Rich 2004, pg. 21) 

Realizing that the thorium data are not measurements, but are the best values that 

the TBD technical staff (Dolan 1988) could reconstruct on the basis of available 

records, recollections of professional engineers, and best estimates on the basis of 

process knowledge, this information represents the best available. From these 

data estimates it is clear that thorium represents less than about 5% of the total 

emissions from the plant processes and that processing occurred during fewer 

years. In addition, these emissions give some qualitative indication of the 

estimated availability of the materials to the airborne pathway. Because the 

contamination controls for thorium processing (ventilation, etc.) were 

the same or equivalent to those for the uranium processes, certain assumptions in 

relation to contaminants in the work place apply to both processes. (Rich 2004, 

pg. 21) 

The data from the report (Dolan 1988) indicates that just the Pilot Plant, Plant 8, 

and Plant 9 processed thorium. A single air sampling data sheet was found that 

indicated a thorium equipment repair operation in Plant 4 during which there 

were air activity concentrations above MAC. Therefore, the three plants 
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mentioned should be considered the primary processing sites, although there is 

some evidence that a few isolated thorium operations occurred in other locations. 

Based on evaluation of the available information, dose reconstructors should 

assume thorium exposure for any employee whose records establish work, and 

therefore exposure potential, primarily in the Pilot Plant from 1964 

to 1979, in Plant 9 in 1954 or 1955, or in Plant 8 in 1966, 1969, 1970, or 1971. 

(pg. 22) 

Some thorium as a contaminant follows the uranium streams through the plant in 

trace quantities but will constitute <1% of the thorium default assumptions below. 

In vivo counting was performed on the workers in the more likely exposed groups 

at least once each year. There is some evidence of urine analyses for thorium in 

claimant files as early as 1955, but to date no information has been found 

regarding how to interpret it. Although urinalysis can offer some information 

regarding thorium intake, it is not the preferred bioassay technique, since the 

material is is predominantly insoluble. Fecal sampling and in vivo analyses are 

the preferred default. This is a difficult default to derive with any degree of 

technical basis. (Rich 2004, pg. 22) 

There was primarily gross alpha and some gross beta air monitoring during 

thorium operations for the purpose of controlling worker exposures to below 

MAC levels. A few in vitro analyses for thorium were discovered primarily in 

claimant file records; only a few in vivo analyses were found. The thorium results 

are questionable because of the lack of information for readily interpreting them 

(e.g., there is no information regarding the in vitro separation method or counting 

procedure/equipment, nor is there information regarding the assumptions made to 

derive the in vivo results). (pg. 22) 

3 - Limited operation times and smaller volumes and mass (which also would 

presuppose a more effective ventilation confinement) reduced the exposure 

potential, all of which would result in an assumption for limited periods of higherlevel contamination. 

4 - The MAC of 100 dpm m-3 (4.5 × 10-11 μCi cm-3) is 20 to 100 times larger 

than the current derived air concentrations for 232Th. 

Based upon the above information and assumptions, the recommended claimantfavorable default exposure approach to assign thorium intakes is to assume: 

• An intake for an exposure period of 100 hours per year at an assumed exposure 

of 10 MAC is judged adequate to account for the higher levels of exposure 

indicated by air sampling, since few samples above 10 MAC were reported and 

these primarily represented short term maximized sampling (based upon 

descriptions on the sample sheets). Also typical and more extensive uranium air 

sample data demonstrate that 10 MAC is a reasonable assumption. 
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• No respiratory protection factor, although not wearing respirators when air 

concentrations were above MAC represented procedural violations. This 

violation was known to have occurred and was not unusual. 

• An intake for an exposure period of 500 hours per year at an average air 

activity of 0.1 MAC during normal operations using these assumptions, the 

claimant-favorable assumption would be: 

500 hr × 0.1 MAC + 100 hr × 10 MAC = 1050 MAC-hr exposure (5-1) 

This exposure results in an intake of about 60 nCi per year. Therefore, in the 

absence of monitoring data a claimant-favorable default intake is 30 nCi/y (82 

pCi/d) each of Th-232 and Th-228 (the alpha emitting isotopes detected on the air 

samples) plus a 60 nCi/y (164 pCi/d) intake of Ra-228. The Ac-228 beta emitter 

adds about 0.1% to the effective dose and therefore can be ignored in the dose 

calculations. (Rich 2004, pg. 23) 

Assuming that those workers receiving the highest weekly doses were those 

workers who would also be the most likely to be in the highest particulate air 

activity, the external dose limits would restrict exposure times to approximately 3 

months per year, i.e., (4000 mrem/312 mrem wk-1) (4.3 wks/month)-1 = 3 

months. This assumes that doses above 4 rem would not be administratively 

planned to avoid exceeding the 5 rem/year limit. In any event external dose limits 

would restrict personnel exposure to the recorded maximum air activity to 

approximately 3 months per year. (Rich 2004, pg. 27) 

From a single radon sample data sheet on which the analyses of two samples 

were recorded on 10/29/53, the higher of the two samples indicated a result of 

230 pCi/L radon gas, which verifies the logical assumption that radon gas was 

released as the drum lids were removed. In addition to the default particulate 

intake (determined as previously stated), a conservative/bounding analysis of 

possible radon plus daughter product exposures can be derived: 

• Assume 230 pCi/L (2.3 WL) with 100% daughter product equilibrium for 1304 

hrs. (163 day x 8 hr/day)/ 74 months of the dumping operations = 17.6 average 

hours/month exposure. Then 2.3 WL 17.6/167(the fraction of a full working 

month) x 12 months = 2.9 WLM exposure per year. 

This assumed bounding exposure to radon plus daughter products would be in 

addition to any assigned exposure that may be derived from the Section 4, 

Occupational Environmental Dose calculations. (Rich 2004, pg. 27) 

It is evident that these estimates are based upon assumptions that are 

cumulatively conservative, claimant favorable, and establish an upper bound of 

intake for workers involved in the transfer operation of the 13,000 barrels of the 

stored MCW raffinates to the K-65 silos. Calculations of internal intakes 

resulting from exposures to the raffinate dusts generated during dumping 
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operations should be used only for claimants for whom a work history on this 

specific project can be established. An examination of external penetrating 

radiation dose for workers who were known to have worked with and handled 

these drums of raffinate wastes show significant (several 100 mrem per week) 

penetrating dose accumulation. Therefore a criteria to determine and/or verify 

that a worker had indeed been exposed to internal intake from raffinate dusts 

would be a record of penetrating external dose, i.e., no detectable dose would 

clearly indicate little direct contact or work with the barrels of waste. (Rich 

2004, pg. 27) 

For internal dose reconstruction of workers in the Silos 1 & 2 areas in the 

absence of specific in vitro and/or in vivo data, a maximization approach, based 

upon air sample results, is proposed: 

• assume a worker was assigned to all of the dumping operations during 6 weeks 

of each of the years in the 6 year period; 

• maximum dumping rate, and hence maximum exposure rate, 80 drums/day; and 

• 100 MAC (4.5 E-9 uCi/ml alpha analysis) exposure levels with no respiratory 

protection at 9.6 E+06 cm3 per day breathing rate. 

The calculation is: 4.5 E-9 μCi/cc × 9.6E+06 cc/day × 30 days = 1.3 μCi intake 

per year for the period of July 1953 to September 1958. The contaminants are 

assumed to be composed of Table 5-16 Isotopes in their indicated activity 

fractions. For purposes of dose reconstruction, it should be noted that the MAC 

air concentrations were determined by gross alpha counts, requiring the addition 

of the appropriate percentage of the beta emitting isotopes. (Rich 2004, pg. 27) 

Assume 230 pCi/L (2.3 WL) with 100% daughter product equilibrium for 

1304 hrs. (163 day x 8 hr/day)/ 74 months of the dumping operations = 17.6 

average hours/month exposure. Then 2.3 WL 17.6/167(the fraction of a full 

working month) x 12 months = 2.9 WLM exposure per year. 

This assumed bounding exposure to radon plus daughter products would be in 

addition to any assigned exposure that may be derived from the Section 4, 

Occupational Environmental Dose calculations. (Rich 2004, pg. 27) 

According to (Rich 2004, Table 5-19, pg. 30), when doing a fecal analysis using 

fluorophotometry, if the MDL is unknown, the dose reconstructor should “assume environmental 

levels of 2 μg per sample. 

Industrial Hygiene & Radiation Department Internal Deposition Action Levels 

procedures from about 1970 indicate actions related to the determination of 

percent maximum permissible lung burden to either uranium or thorium. 

Uranium-235 was detected primarily by the emission of its 186 keV photon. 
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Uranium-238 was calculated from measurement of the Th-234 progeny assumed 

to be in equilibrium with the U-238. Thorium-232 and Th-228 activities were 

determined based on equilibrium assumptions and detection of their progeny, 

most likely Ac-228 for Th-232, but Pb-212 may have been used for assessment of 

both Th isotopes. Thorium-230 is not readily detectable by in vivo measurements. 

There appeared to be no attempt to detect TRU contaminants with the MIVRML. 

In fact, the only determination made with the mobile van was a quantification of 

the uranium lung burden in micrograms of uranium, with the assumption of 1% 

enrichment and of occasional thorium lung burdens as indicated by some claim 

records. (Rich 2004, pg. 35) [This has been used later] 

The results from the MIVRML were calibrated in μCi of 235U and reported in mg 

of uranium in the lung, which was translated to MPLB based on the assumed 

enrichment (generally 1%). The percent body burden was then multiplied by 15 

rem to obtain the assigned annual dose in rem. (Rich 2004, pg. 35) 

The recorded MDA reported in Scott et al. (1969, p. 169). Values reported in a 

claimant record were below 10% MPLB, which implies an MDA of less than 1 

nCi total 1% (assumed) U-235. (Rich 2004, pg. 36, footnote to Table 5-26). 

The following general assumptions were noted in the Y-12 Occupational Internal Dose TBD 

(Rich and Chew 2006). Enrichment assumptions at Y-12 for dose assessment for various 

analytical techniques were provided in Table 5-1 (Rich and Chew 2006, pg. 13) 

Based on the changing workplace conditions after partial restart in 1998, the 

default assumptions were modified to return to chronic exposure but to use type S 

solubility. (Rich and Chew 2006, pg. 12) 

Before stand-down, the Y-12 program default modeling assumption was class Q 

(90% Super-W, 10% Y). During the stand-down, the Y-12 program default 

assumption of chronic exposure was modified to assume acute exposures 

occurring at the midpoint of a quarterly sampling frequency. (Rich and Chew 

2006, pg. 11) 

Exposure to type M material from 1948 to July 1998 appears to be the more likely 

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

likely. However, the absorption type can be based on the monitoring data, 

claimant-favorable assumptions, or both. (Rich and Chew 2006, pg. 12) 

For lung counts a combination of the information in the Type Analysis and 

Material Type (see Section 5.A.1.3 for details) reporting fields can be used to 

determine whether the count was believed to be due to NU or DU. For records 

through 1971, a Type Analysis of 1 with a Material Type of 2 or 3 indicates DU, 

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

of 4 with a Material Type of 2 or 3 indicates DU, while a Material Type of 7 

indicates NU. (Rich and Chew 2006, pg. 13) 
 

 

 

 

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

nature. For example, in 1958, Patterson stated, “...our interpretation of 

urinalysis results and our assignment of internal dose assume an exposure under 

equilibrium conditions of intake and elimination” (Patterson 1958, p. 58). For 

the stand-down period from September 1994 until August 1998, acute exposures 

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

complete equilibrium is not expected in modern internal dosimetry models, the 

presumption of chronic exposure conditions for uranium remains in place as 

noted in a 2003 report: “The most likely exposure potential for uranium work at 

the Y-12 Complex is chronic in nature” (BWXT Y-12 2003, p. 107). (Rich and 

Chew 2006, pg. 10) 

Reference to Table 5-8 with relationship to the plutonium contaminant, the 

recommended isotope Pu-239 should be assumed for Oralloy only, since Oralloy 

used at Y-12 did not come directly from the primary RU generating sites, i.e., it 

was enriched at the GDP facilities in which Pu-239 was dominant compared to 

Pu-238. (Rich and Chew 2006, pg. 19) 

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

1965. However, given the limitations of the rate method of estimating daily urine 

volumes, uncertainty in the excretion volume is expected to add substantially to 

the uncertainty associated with the detection limit of a single measurement. 

Because recoveries were based on batch rather than individual measurements, 

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

of a single measurement. (Rich and Chew 2006, pg. 23) 

The results in the claim files are assumed to be in units of activity per sample 

unless otherwise indicated (BWXT Y-12 2003, p. 7). (Rich and Chew 2006, 

pg. 24) 

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

alpha spectroscopy began to be used. Fluorometry yielded results in mass units 

(e.g., μg/L), but results were often converted to disintegrations per minute per day 

with the assumed specific activity of 1.55 dpm/μg for NU. The fluorometric 

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

1949, p. 7). (Rich and Chew 2006, pg. 25) 

Almost all of the restrictions that did take place would have occurred on the basis 

of the urinalysis program alone. In those cases in which restriction was based on 

in vivo analysis alone, lung retention times were often observed to be 

considerably longer than the assumed 120-d lung half-life on which the urinalysis 

program was based. In other words, the problem was largely due to the 

inapplicability of the lung model for some materials and individuals, rather than 

a failure of the urinalysis program. (Rich and Chew 2006, pg. 26) 

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

made about solubility and uranium activity ratios. (Rich and Chew 2006, pg. 27) 
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The beta activity of the tritium is then measured by liquid scintillation counting.... 

Exposure should be assumed to be due to HTO. (Rich and Chew 2006, pg. 27) 

Assume that the ratio of 234U to 238U is 1:1 for dietary uranium. (Rich and Chew 

2006, pg. 29) 

Due to the uncertainty about actual methods for determining the lung counting 

detection limits, for dose reconstruction purposes and based on review of the Y-12 

data, the 235U and 238U lung count detection thresholds are assumed to be 130 μg 

and 13.5 mg, respectively, through 1990. Beginning in 1991, the critical level or 

MDA is supplied with the individual results. (Rich and Chew 2006, pg. 31) 

No description of the conversion count rate to activity to mass has been found. 

The analyte reported was based on the area in which the employee worked. 

Individuals working in NU or DU areas had results reported as 238U, and workers 

in enriched areas had results reported as 235U. Claimant-favorable assumptions 

should be based on conversions of 93% enrichment for 235U and natural isotopic 

abundances for 238U. (Rich and Chew 2006, pg. 32) 

This means that the maximum dose conversion factor per milligram of 232Th 

would be less than that for 232Th in full equilibrium with its progeny. However, 

unless specific information is available in the claimants’ data, the dose 

reconstructor must make claimant-favorable assumptions. The thorium 

sensitivity varied due to the dependence of this technique on the ratios of 232Th 

and 228Ra to the progeny being measured. (Rich and Chew 2006, pg. 33) 

Exposure to type M material from 1948 to June 1998 appears to be the most likely 

absorption type. After June 1998, exposure to absorption type S material is more 

likely. However, the absorption type can be based on the monitoring data or 

claimant-favorable assumptions. (Rich and Chew 2006, pg. 46) 

For dose reconstruction purposes and based on review of the Y-12 data, the 235U 

and 238U lung count detection thresholds are assumed to be 130 μg and 13.5 mg, 

respectively, through 1990. Beginning in 1991, the critical level or MDA is 

supplied with the individual results. (Rich and Chew 2006, pg. 47) 

The Paducah Occupational Internal Dose TBD (Berger 2004, pg. 6) describes the radionuclides 

of concern at PGDP in the following statement: 

The mission of PGDP was to enrich uranium in the form of UF6 (for use in 

domestic and foreign commercial power reactors) from roughly 0.7% 235U 

(natural enrichment) to 2.5% 235U (DOE 2000, p. 6). In addition, other 

compounds of uranium were present throughout the plant’s history, including 
 

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UO2F2, UF4, and UO3. The primary radionuclides of concern for the plant are

238U, 235U, and 234U. The progeny of dosimetric interest for these radionuclides 

includes 230Th and 234mPa.... (Berger 2004, pg. 6) 

At Fernald, the Occupational Internal Dose TBD (Rich 2004) points summarize three main 

radionuclides of concern. 

This material remains on the Fernald site in the K-65 Silos and represents a 

continuing internal exposure potential (for any operations in which direct contact 

with these residues is required) from the unusually high concentrations of 226Ra 

and its progeny, 210Pb and 210Po (these three radionuclides comprise 

approximately 90% of the total activity in the K-65 materials). (Rich 2004, pg. 7) 

The Fernald Occupational Internal Dose TBD goes much farther than the Paducah 

internal dose TBD in providing a detailed list of radionuclides of concern, (Rich 2004, 

Table 5.2). The Y-12 Occupational Internal Dose TBD (Rich and Chew 2006), likewise 

provides a fuller listing of radionuclides of concern. 

The Paducah Occupational Internal Dose TBD (Berger 2004) fails to even mention airborne 

releases. 

The Fernald Occupational Internal Dose TBD has numerous places where airborne release 

potentials are mentioned. 

Plant 6, the Metals Fabrication Plant, and Plant 9 produced metal parts in 

rolling mills and machining lathes. Plant 6 began operations in 1953. Uranium 

metal fires were common, resulting in elevated airborne uranium concentrations. 

(Rich 2004, pg. 9) 

In addition to the routine releases at FEMP, there were frequent “upset” 

conditions (i.e., spills, effluent filter ruptures, etc.) that produced episodic 

airborne radioactivity in the work areas and plant effluents, and were of a 

magnitude that the ventilation systems were unable to contain all of the releases. 

(Rich 2004, pg. 11) 

Before 1989 no TRU analyses for radiological safety were performed on a routine 

basis for either airborne or urine activity, and exposure controls remained based 

on chemical toxicity under the assumption that these controls would be sufficient 

for all the radiological issues (Bassett et al. 1989). Although the alpha activity 

from the TRU alpha emitters would have been collected and detected on the air 

samples, the reported results were all considered to be uranium and compared to 

the MAC. (Rich 2004, pg. 14) 

The information on the data sheets indicates that in spite of the fact that the 

contents of the drums were wet, the operations resulted in significant airborne 

contamination. (Rich 2004, pg. 26) 
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The Y-12 Occupational Internal Dose TBD (Rich and Chew 2006), also mentions airborne 

releases on three different pages. 

The 86-inch Cyclotron started operation on November 11, 1950, and operated 

until the early 1980s. This cyclotron is no longer in operation. Polonium 

isotopes and alpha airborne activity are the mentioned internal dose concerns. 

The 86-inch cyclotron was later used to create neutron-deficient radionuclides as 

a part of the R&D program above. (Rich and Chew 2006, pg. 10) 

If a work area has been determined to have predominantly insoluble airborne 

uranium, it is current practice for workers with a moderate to high exposure 

potential (CEDE greater than 100 mrem) to submit both urine and fecal samples. 

The relative elimination by urinary and fecal pathways is used to determine the 

solubility mixtures for each individual. (Rich and Chew 2006, pg. 28) 

The Paducah Occupational Internal Dose TBD (Berger 2004) highlights the following in regard 

to monitoring for intakes of uranium. 

At the PGDP, monitoring for intakes of uranium, whether in vivo or in vitro, often 

resulted in reports of elemental uranium concentration in urine or the mass of 

elemental uranium in organs or the whole body. However, internal dose 

assessment requires the use of isotopic concentrations as input to the assessment 

process. Therefore, reconstructors should use Table 5-2 to derive the isotopic 

fractions associated with each microgram of uranium reported in an analytical 

result. (Berger 2004, pg. 7) 

Other assumptions related to in vitro analysis in the Occupational Internal Dose TBD (Berger 

2004) include those in vitro measurement frequencies that are noted in Table 5-6 on page 13 and 

in vitro measurement types and detection levels for various periods noted in Table 5-7 on 

page 14. 

At Paducah, whole-body counting was performed using a mobile counter provided by the Y-12 

plant that was sometimes referred to as the Martin Marietta Energy Systems (MMES) Counter. 

Table 5-10 on page 18 of the Paducah Occupational Internal Dose TBD (Berger 2004) lists 

general information about the detection capabilities of this counting system for various periods 

and provides MDAs for various operational periods at PGDP. In the Fernald Occupational 

Internal Dose TBD (Rich 2004, pg. 34), Table 5-22 provides a similar table of MDA which 

shows the MDA for non-uranium radionuclides was 0.1 pCi/sample throughout most of the 

years. Uranium in vivo MDAs for the lung at Fernald are provided in Table 5-26 on page 36. 

Whereas, the MDA at Paducah was 4 mg for U-238 from 1968 to 1980 (Berger 2004, pg. 18, 

Table 5-10), the MDA for U-238 from 1989 to 2001 at Fernald was 7.4 mg (Rich 2004, pg. 36, 

Table 5-26). The lung counting detection levels at Y-12 using two 9” × 2.5” NaI detectors 

varied from 13.5 mg in 1959 down to 4.5 mg when the HP germanium detectors were installed 

(Rich and Chew 2006, Table 5-12, pg. 32). 
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In the Paducah Occupational Internal Dose TBD (Berger 2004, pp. 5 and 8), only two references 

to potential thorium exposure are noted. These involve processing small quantities of transuranic 

(TRU) elements, primarily neptunium and plutonium present in various workplaces, as well as 

the thorium and protactinium progeny of uranium and the fission product 99Tc. At Fernald, 

(Rich 2004, pg. 8) it mentions that more than 70% of the thorium that was handled and processed 

occurred during the period from 1964 to 1979 in the Pilot Plant. There are numerous discussions 

of thorium in the Fernald Occupational Internal Dose TBD. In Table 5A-1 on page 36 of the 

ORNL Occupational Internal Dose TBD (Bollenbacher et al. 2006), data used to determine 

historical MDAs for radionuclides in urine and feces are listed. Thorium is one of the listed 

radionuclides. In the LANL internal dose TBD (Argall 2004), there is more discussion regarding 

the isotopic mixture for thorium based on professional judgment. In the case of a maximizing 

approach, the LANL TBD assumes Th-230. In the LANL internal dose TBD, it advised the dose 

reconstructor that the absorption class (M or S) should be selected based on the expected 

compound or matrix (pg. 33). 

The use of IMBA in the Paducah Occupational Internal Dose TBD (Berger 2004, pg. 7) is only 

mentioned as the source of the isotopic fractions for various enrichment percentages in 

Table 5-2. In the Fernald Occupational Internal Dose TBD (Rich 2004, pg. 10) it is noted that 

the mass percentages, relative activities in pCi/μg, and the total pCi/μg values are based upon 

IMBA NIOSH default values that are provided in Table 5.3 (Rich 2004, pg. 10). Thus, the 

Fernald Occupational Internal Dose TBD more fully explains the associated activities 

determined by IMBA for different categories at Fernald. In the LANL internal dose TBD, the 

isotopic composition for uranium is defaulted to the values in the Integrated Modules for 

Bioassay Analysis IMBA NIOSH Phase I database USDOE Version 1.0.42, Table 5-11 (depleted 

uranium), or Table 5-12 (enriched uranium) (Argall 2004, pp. 25–26). 

Uranium background determination is not discussed in the Paducah Occupational Internal Dose 

TBD. The Fernald Occupational Internal Dose TBD (Rich 2004, pg. 28) does mention 

background regarding minimal occupational internal exposure above normal expected 

radon/thoron backgrounds from the Silos source and mentions that a default of 5% of 0.4 WL 

could be added to plant exposures. It is pointed out that this equals an exposure of 0.24 WLM 

per year. The Oak Ridge National Laboratory Internal Dosimetry Program Technical Basis 

Document (McLaughlin 2002) documented the ORNL assumptions on the handling of uranium 

exposure: 

....A urinary uranium background study was conducted in the mid 1990's using 

non-occupationally exposed employees to quantify the range of typical 

background uranium excretion. Based upon the results of that study, a 

discrimination level (set at the 99th percentile level) of 0.14 dpm/day was 

established for both U-234 and U-238 to differentiate between environmental and 

occupational exposure to uranium. A value of 0.25 dpm/day is applied to total 

uranium results. (McLaughlin 2002) (Also cited in the internal dose TBD on page 

18) 

The Paducah Occupational Internal Dose TBD (Berger 2004) is very thorough in providing 

solubility classifications and assumed particle size (μm AMAD) for a number of radionuclides 
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by facility or building in Table 5.5 on pages 8–12. A general statement is also made about 

particle size as well (Berger 2004, pp 7). 

At the PGDP, and unless site-specific information is available, the particle 

size is assumed to be 5 μm aerodynamic median activity diameter. (Berger 2004, 

pg. 7) 

It is stated in the Fernald Occupational Internal Dose TBD (Rich 2004, pg. 19) that “Table 5-14 

lists a summary of the thorium plant processes with chemical forms and solubility types.” 

However, a review of Table 5.14 on page 20 only lists production years and chemical form, 

leaving the dose reconstructor to try to attempt to assign the proper solubility type based on the 

chemical form provided. At Y-12, the following statement was made in the Y-12 Occupational 

Internal Dose TBD. 

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

tended to be lower, for materials processed at higher temperatures.... While the 

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

more important to note that, for the vast majority of individuals, lung clearance 

took place in approximate accordance with the ICRP Publication 2 (1960) 

insoluble model, which fits within the current type M framework. (Rich and Chew 

2006, pg. 10) 

The predominant material encountered after partial restart in August 1998 is 

uranium oxide, which fecal sampling has shown to be more consistent with 100% 

type S material. Based on the changing workplace conditions after partial restart 

in 1998, the default assumptions were modified to return to chronic exposure but 

to use type S solubility. (Rich and Chew 2006, pg. 11) 

For a workplace as varied as Y-12, it is clear that no single solubility or particle 

size would apply to all workers. Furthermore, accurate assignment of the 

uranium lung clearance type to a given bioassay result was considered virtually 

impossible because of uncertainties about chemical form and limitations of the 

personnel-tracking system (Barber and Forrest 1995, p. 669). (Rich and Chew 

2006, pg. 12) 

In the Paducah Occupational Internal Dose TBD (Berger 2004, pg. 15), Table 5-8 indicates that 

alpha spectrometry was used as the analytical method for urine bioassay samples, and a detection 

level of 0.27 pCi/L was given for use with Th-228, Th-232, U-234, U-235, U-238, Pu-238, 

Pu-239, Pu-240 and Am-241. 

In the Fernald Occupational Internal Dose TBD, the following explanation was provided on the 

spectrometer used at Fernald. 

The method currently being used at Fernald for urinalysis is inductively coupled 

plasma mass spectrometry (ICP-MS) which has an a priori MDA of 0.15 μg L-1. 

(Rich 2004, pg. 32) 
 

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At Fernald, there also is no mention of evaluation dose for strontium exposure but the 

Occupational Internal Dose TBD did provide the dose reconstructor with the following 

information on polonium and cesium. 

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

and thorium were quantified so that their effects on the spectra could be taken 

into account: 

• 40K, 137Cs 

Other radionuclides addressed included: 

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

• 210Po (Rich and Chew 2006, pg. 13) 

In regard to the potential for Cs-137 exposure at PGDP, the Occupational Internal Dose TBD 

provides the following statement. 

On occasion, in vivo measurement results included 137Cs. However, those PGDP 

workers could have had body burdens of 137Cs from nonoccupational sources 

(e.g., fallout and consumption of specific foodstuffs). There is no evidence of 

occupational intakes of 137Cs at the PGDP, thus no dose of record should be 

associated with these measurement result, if any. (Berger 2004, pg. 19) 

The Fernald Occupational Internal Dose TBD provides no information of possible cesium 

exposure to FEMP workers. 

The Y-12 Occupational Internal Dose TBD does discuss the contribution of dose from Cs-137. 

The 137Cs contribution to the regions of interest changed over time as fallout 

levels varied, requiring frequent updates to the control subject spectrum library. 

(Rich and Chew 2006, pg. 31) 

Therefore, it is concluded that the primary internal radiation exposure to 

Y-12 workers was from uranium. However, the internal dosimetry program has 

included limited monitoring for 137Cs, 99Tc, thorium, plutonium, 228Ac, and 

tritium, among other radionuclides. There are difficulties in interpreting these 

measurement data that cannot be resolved satisfactorily at this time. These issues 

should not have a large impact on the worker’s internal dose because exposure to 

uranium is the source of the greatest part of the internal dose. (Rich and Chew 

2006, pg. 46) 

The Paducah Occupational Internal Dose TBD discusses thorium exposure only in one place. 

Over the years, workers at PGDP handled mainly UF6 and slightly oxidized forms 

of uranium. The facility processed both virgin feed material and recycled or 

reprocessed reactor fuel to enrichments of up to 5% (by weight) of 235U in the 
 

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final product.1 This processing involved small quantities of transuranic (TRU) 

elements, primarily neptunium and plutonium, present in various workplaces, as 

well as the thorium and protactinium progeny of uranium and the fission product 

99Tc. (Berger 2004, pg. 5) 

The Fernald Occupational Internal Dose TBD provides guidance on assumption for 

environmental levels of thorium and its contribution to internal dose. 

For the purposes of dose reconstruction, effective equilibrium is a logical, 

claimant-favorable assumption because thorium was present from the earliest 

times and was stored at FEMP after the industry need for thorium products was 

past. However, for purposes of perspective Figure 5-1 presents thorium 

equilibrium growth curves. (Rich 2004, pg. 19). 

Realizing that the thorium data are not measurements, but are the best values that 

the TBD technical staff (Dolan 1988) could reconstruct on the basis of available 

records, recollections of professional engineers, and best estimates on the basis of 

process knowledge, this information represents the best available. From these 

data estimates it is clear that thorium represents less than about 5% of the total 

emissions from the plant processes and that processing occurred during fewer 

years. In addition, these emissions give some qualitative indication of the 

estimated availability of the materials to the airborne pathway. Because the 

contamination controls for thorium processing (ventilation, etc.) were the same or 

equivalent to those for the uranium processes, certain assumptions in relation to 

contaminants in the work place apply to both processes. (Rich 2004, pg. 21) 

The data from the report (Dolan 1988) indicates that just the Pilot Plant, Plant 8, and 

Plant 9 processed thorium. A single air sampling data sheet was found that indicated a 

thorium equipment repair operation in Plant 4 during which there were air activity 

concentrations above MAC. Therefore, the three plants mentioned should be considered 

the primary processing sites, although there is some evidence that a few isolated thorium 

operations occurred in other locations. Based on evaluation of the available 

information, dose reconstructors should assume thorium exposure for any employee 

whose records establish work, and therefore exposure potential, primarily in the Pilot 

Plant from 1964 to 1979, in Plant 9 in 1954 or 1955, or in Plant 8 in 1966, 1969, 1970, 

or 1971. (p. 22) 

Some thorium as a contaminant follows the uranium streams through the plant in 

trace quantities but will constitute <1% of the thorium default assumptions below. 

In vivo counting was performed on the workers in the more likely exposed groups 

at least once each year. There is some evidence of urine analyses for thorium in 

claimant files as early as 1955, but to date no information has been found 

regarding how to interpret it. Although urinalysis can offer some information 

regarding thorium intake, it is not the preferred bioassay technique, since the 

material is predominantly insoluble. Fecal sampling and in vivo analyses are the 
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preferred default. This is a difficult default to derive with any degree of technical 

basis...( Rich 2004, pg. 22). 

According to (Rich 2004, Table 5-19, pg. 30), when doing a fecal analysis using 

fluorophotometry, if the MDL is unknown, the dose reconstructor should “assume 

environmental levels of 2 μg per sample. 

Industrial Hygiene & Radiation Department Internal Deposition Action Levels 

procedures from about 1970 indicate actions related to the determination of 

percent maximum permissible lung burden to either uranium or thorium. 

Uranium-235 was detected primarily by the emission of its 186 keV photon. 

Uranium-238 was calculated from measurement of the Th-234 progeny assumed 

to be in equilibrium with the U-238. Thorium-232 and Th-228 activities were 

determined based on equilibrium assumptions and detection of their progeny, 

most likely Ac-228 for Th-232, but Pb-212 may have been used for assessment of 

both Th isotopes. Thorium-230 is not readily detectable by in vivo measurements. 

There appeared to be no attempt to detect TRU contaminants with the MIVRML. 

In fact, the only determination made with the mobile van was a quantification of 

the uranium lung burden in micrograms of uranium, with the assumption of 1% 

enrichment and of occasional thorium lung burdens as indicated by some claim 

records. (Rich 2004, pg. 35) 

In regard to the potential dose for the actinides, the Y-12 Occupational Internal Dose TBD, 

Table 5-9 on page 24 lists the analytical laboratory LD values used at Y-12 from 1988 for the 

radionuclides: Am-241, Th-228, Th-232, Np-237. 

The in vivo lung count was the only monitoring technique for monitoring thorium 

exposure in the body during the plant’s first decades. Thorium lung activity was 

inferred from 228Ac and/or 212Pb lung activity. Thorium lung counting was conducted 

from 1958 to 1984 with routine lung counts, scheduled at approximately six-month 

intervals, starting in 1961. (Rich and Chew 2006, pg. 32) 

Dose Reconstruction Assumptions for Occupational External Dose 

The Paducah Occupational External Dose TBD (Turner and Thomas 2006) lists the following 

assumptions for use by the dose reconstructor. 

Summary of reasonable but claimant favorable assumptions: 

1. Neutron to photon ratio developed using the average midline and bottom 

photon dose rates. 

2. Neutron to photon ratio developed using the low enriched uranium neutron 

dose rate 

3. Assumption that all recorded and missed photon dose was the result of 

exposure to enriched UF4 material 

4. Assumption that all of the neutron energies are between 0.1–2 MeV 
 

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5. Assumption that neutron dosimeter used a quality factor of 10 to convert 

rad to rem. (Turner and Thomas 2006, pg. 24) 

...this TBD has assumed that the TLD was used in a wrist dosimeter 

configuration, and that dose calculations involved use of a modifying factor to 

provide some estimate of actual extremity exposure. The sampling of records 

reviewed showed that while extremity doses were often near limits, the imposed 

whole-body restrictions limited worker extremity exposure to less than the 

extremity limit. It was concluded that the dose of record is the best 

to use for reconstruction of the extremity dose. (Turner and Thomas 2006, pg. 11) 

The use of the mrep unit is somewhat unique to FEMP because it declined in use 

after the 1950s. There are few references, including the Radiological Health 

Handbook (PHS 1970), so this TBD assumes that a rep is approximately 93 

ergs/g of tissue. Because FEMP used this term interchangeably with rad (100 

ergs/g of any receptor), there is a further inherent conservativeness of 

approximately 7%. (Turner and Thomas 2006, pg. 13) 

It is reasonable to assume that the early versions of the FEMP dosimeters reacted 

similarly to the ORNL dosimeter system, given that FEMP used the ORNL system 

until changing to a commercial system in 1985. (Turner and Thomas 2006, 

pg. 13) 

Results of tests of FEMP dosimeters used during the 1960s (Heatherton 1960) 

included the conclusion that the half-value thickness of absorption of UX-2 

(234Pa) beta energy was approximately 110 mg/cm2. It was determined that “the 

combined dose rate from the surface of uranium metal in equilibrium with its two 

daughters, UX-1 (234Th) and UX-2 (234Pa), is about 240 mrad/hr.” It was also 

determined that approximately 95% of the surface dose rate, or approximately 

228 mrad/hr, originated from the UX-2 in the metal. The processing of the metal 

resulted in separation of uranium daughter products, which produced much 

higher dose rates in portions of the product, process equipment, and byproducts. 

The reason for this increase in dose rates is the loss of self-shielding afforded by 

the mass of the in-process uranium. Further studies involved the absorption 

afforded by such materials as film wrapping paper, polyethylene, cardboard, and 

Lucite. These materials were used because they were assumed to be nearly tissue 

equivalent. (Turner and Thomas 2006, pg. 14) 

While it was not explicitly stated in the documentation (Heatherton 1960), it is 

assumed that the dosimeter in use at FEMP was the ORNL version and that the 

film used was a DuPont type (Johnson 1963). There is some general data of 

results between film and TLD dosimeters dated November and December 1982. 

After review of these data, the conclusion is that the two types of dosimeters did 

not agree and that the ratio of film to TLD varied with the location of the 

exposure. In some instances the ratio was greater than 1, and in others it was 

less than 1. Most often film results provide higher dose estimates than TLD 

results, which support the conclusion that early film dosimeter results are 
 

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claimant-favorable. However, TLD results have been documented to be more 

accurate than film dosimeters, and provide a more representative measure of the 

true exposures (this is particularly true when measuring beta and low-energy 

photons). (Turner and Thomas 2006, pg. 14) 

The most claimant favorable neutron energy group (i.e., the neutron energy group 

with the largest REF) is the 0.1 – 2.0 MeV energy group. Since the neutrons at 

Fernald are assumed to be generated (born) in this energy group and although 

some fraction will scatter to a lower (less claimant favorable) energy groups, all 

of the neutron dose should be assumed to result from the 0.1 – 2 MeV energy 

group. In the absence of data, this is a reasonable and claimant favorable 

assumption. The default neutron dose fractions are given in Table 6-11. (Turner 

and Thomas 2006, pg. 22) 

It can be assumed with some certainty that there have been missed doses in the 

recorded doses for FEMP workers. This could have resulted when a dosimeter 

was lost or a worker was not monitored, or a zero was entered because the 

dosimeter result was less than the MDL. Various methods were used to estimate 

lost dosimeter results such as using coworker results, products of time and dose 

rate, or previously recorded results for similar work. All of these required 

considerable review and examination. Missed dose from MDLs is especially 

important when there were short exchange periods, generally through the 1950s 

and 1960s. That period also had higher MDLs. The recommended procedure for 

missed dose is to assign with a log-normal distribution, with zero as a minimum 

value, (LOD/2 * the number of zero measurements) as the central tendency and 

(LOD * the number of zero measurements) as the upper 95% estimate. This 

procedure is applicable to both Hs(0.07) and Hp(10) since the same dosimeter 

was used for both until the introduction of a TLD finger dosimeter. The MDL per 

period and exchange frequencies along with the product (LOD * exchanges) 

are listed in Table 6-13. (Turner and Thomas 2006, pg. 23) 

The Fernald Occupational External Dose TBD (Faust 2004) listed the following assumptions for 

use by the dose reconstructors. 

General assumptions were listed together in one location in the Fernald Occupational External 

Dose TBD for use by the dose reconstructor, as was the case with the Paducah Occupational 

External Dose TBD. 

Summary of reasonable but claimant favorable assumptions: 

1. Neutron to photon ratio developed using the average midline and bottom 

photon dose rates. 

2. Neutron to photon ratio developed using the low enriched uranium neutron 

dose rate 

3. Assumption that all recorded and missed photon dose was the result of 

exposure to enriched UF4 material 

4. Assumption that all of the neutron energies are between 0.1–2 MeV 
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5. Assumption that neutron dosimeter used a quality factor of 10 to convert 

rad to rem. (Faust 2004, pg. 24) 

More specific additional Fernald Occupational External Dose TBD assumptions were also 

provided 

TLDs were introduced in or around 1978 or 1979, but only on an experimental 

basis. An exception was the extremity dosimeter program, which introduced 

TLDs in 1977; however, no data were found in the literature that described these 

TLDs other than “they were the Teledyne Teflon impregnated with calcium 

sulfate type” (Dugan 1981). Therefore, this TBD has assumed that the TLD was 

used in a wrist dosimeter configuration, and that dose calculations involved use 

of a modifying factor to provide some estimate of actual extremity exposure. 

(Faust 2004, pg. 11) 

This TBD assumes that when FEMP began the practice of using bags for 

contamination control, attendant calibration procedures were revised to 

accommodate the new practice. Small changes in film density can mean large 

changes in recorded exposure. Therefore, it is desirable to calibrate 

dosimeters using the same radiological sources and energies to which worker 

dosimeters will be exposed; this practice was implemented at FEMP. (Faust 

2004, pg. 13) 

...this TBD assumes that a rep is approximately 93 ergs/g of tissue. Because 

FEMP used this term interchangeably with rad (100 ergs/g of any receptor), there 

is a further inherent conservativeness of approximately 7%. (Faust 2004, pg. 13) 

It is reasonable to assume that the early versions of the FEMP dosimeters reacted 

similarly to the ORNL dosimeter system, given that FEMP used the ORNL system 

until changing to a commercial system in 1985. (Faust 2004, pg. 13) 

Results of tests of FEMP dosimeters used during the 1960s (Heatherton 1960) 

included the conclusion that the half-value thickness of absorption of UX-2 

(

234Pa) beta energy was approximately 110 mg/cm2 . It was determined that “the 

combined dose rate from the surface of uranium metal in equilibrium with its two 

daughters, UX-1 (234Th) and UX-2 (234Pa), is about 240 mrad/hr.” It was also 

determined that approximately 95% of the surface dose rate, or approximately 

228 mrad/hr, originated from the UX-2 in the metal. The processing of the metal 

resulted in separation of uranium daughter products, which produced much 

higher dose rates in portions of the product, process equipment, and 

byproducts. The reason for this increase in dose rates is the loss of self-shielding 

afforded by the mass of the in-process uranium. Further studies involved the 

absorption afforded by such materials as film wrapping paper, polyethylene, 

cardboard, and Lucite. These materials were used because they were assumed to 

be nearly tissue equivalent. It was determined that the half value thickness for 

tissue was approximately 110 mg/cm2

 and, therefore, the dose to the eyes or 
 

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gonads was approximately 15% of the skin dose. It was also determined that 

coveralls worn by workers (about 30 mg/cm2) reduced uranium beta exposure to 

the skin by approximately 20%. Figure 6-6, at the end of this TBD, summarizes 

these data. (Faust 2004, pg. 14) 

While it was not explicitly stated in the documentation (Heatherton 1960), it is 

assumed that the dosimeter in use at FEMP was the ORNL version and that the 

film used was a DuPont type (Johnson 1963). (Faust 2004, pg. 14) 

Since the neutrons at Fernald are assumed to be generated (born) in this energy 

group and although some fraction will scatter to a lower (less claimant favorable) 

energy groups, all of the neutron dose should be assumed to result from the 0.1 – 

2 MeV energy group. In the absence of data, this is a reasonable and claimant 

favorable assumption. The default neutron dose fractions are given in 

Table 6-11. (Faust 2004, pg. 22) 

The probability of causation calculations in IREP uses a Radiation Effectiveness 

Factor (REF) to estimate the probability of causation. The most claimant 

favorable neutron energy group (i.e., the neutron energy group with the largest 

REF) is the 0.1 – 2.0 MeV energy group. Since the neutrons at Fernald are 

assumed to be generated (born) in this energy group and although some fraction 

will scatter to a lower (less claimant favorable) energy groups, all of the neutron 

dose should be assumed to result from the 0.1 – 2 MeV energy group. In the 

absence of data, this is a reasonable and claimant favorable assumption. The 

default neutron dose fractions are given in Table 6-11. (Faust 2004, pg. 22) 

It can be assumed with some certainty that there have been missed doses in the 

recorded doses for FEMP workers. This could have resulted when a dosimeter 

was lost or a worker was not monitored, or a zero was entered because the 

dosimeter result was less than the MDL. Various methods were used to estimate 

lost dosimeter results such as using coworker results, products of time and dose 

rate, or previously recorded results for similar work. All of these required 

considerable review and examination. Missed dose from MDLs is especially 

important when there were short exchange periods, generally through the 1950s 

and 1960s. That period also had higher MDLs. The recommended procedure for 

missed dose is to assign with a log-normal distribution, with zero as a 

minimum value, (LOD/2 * the number of zero measurements) as the central 

tendency and (LOD * the number of zero measurements) as the upper 95% 

estimate. This procedure is applicable to both Hs(0.07) and Hp(10) since the 

same dosimeter was used for both until the introduction of a TLD finger 

dosimeter. The MDL per period and exchange frequencies along with the product 

(LOD * exchanges) are listed in Table 6-13. (Faust 2004, pg. 23) 

The Y-12 Occupational External Dose TBD (Kerr 2006), which has been recently 

revised, has assumptions that are based on a different approach from the approach used in 

the Paducah and Fernald Occupational External Dose TBDs. 
 

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Claimant-favorable assumptions should be made using guidance in Table F-2 for 

beta particles and for photons (x-rays and gamma rays) to assure that dose is not 

underestimated. The values presented in this table are intended to provide a 

reasonable estimate of parameters used to calculate the organ dose without 

significant numerical error for long-term Y-12 workers in the respective facilities. 

There is no direct evidence to select the specific values shown other than 

considerations of the radiation sources and usual work tasks. In those cases 

where there is some doubt in the values, a range of realistic values should be 

selected for comparison and the most claimant-favorable option selected. (Kerr 

2006, pg. 43) 

A detailed discussion of the conversion of measured dose to organ dose 

equivalent is provided in Appendix A of NIOSH (2002). Appendix B of NIOSH 

(2002) contains the appropriate dose conversion factors (DCFs) for each organ, 

radiation type, and energy range based on the type of monitoring performed. In 

some cases, simplifying assumptions are appropriate. (Kerr 2006, pg. 49) 

The two general types of neutron dosimeters that have been used at Y-12 differ 

significantly in their response to neutrons of different energies as illustrated in 

Figure 6.3.2.2-1 (IAEA 1990). An NTA emulsion was included in the same holder 

used for the Y-12 beta/gamma dosimeter until 1980. Between 1980 and 1989, 

there is a serious gap in the neutron dosimetry information for Y-12. It is 

known that Y-12 had become increasingly dependent over the years on ORNL to 

process the NTA films because of the small numbers of neutron-exposed workers 

at Y-12. Thus, the neutron dosimetry at Y-12 is assumed to be the same as that at 

ORNL from 1980 to 1989. During this period, workers at both ORNL and Y-12 

were provided with a two-element TLD dosimeter for beta-particle and photon 

dosimetry. Those ORNL and Y-12 workers who were exposed to neutrons were 

provided with a separate neutron dosimeter. This neutron dosimeter contained 

both an NTA film for measurement of the fast-neutron dose and a TLND for 

measurement of the neutron dose from lower energy neutrons (Gupton 1978; 

Berger and Lane 1985). From 1980 to 1985, the neutron doses to Y-12 workers 

were determined at ORNL using both the NTA and TLND dosimeters as discussed 

by Gupton (1978). From 1986 to 1989, they were determined at ORNL using only 

the TLNDs (Berger and Lane 1985). (Kerr 2006, pp. 18–19) 

This dosimeter was a modification of the film badge dosimeter previously used at 

both ORNL and Y-12. During the switch from film to TLD, the film badge 

dosimeter was modified to hold four TLD chips in a polyethylene mount, 1 mm 

thick. For workers exposed to neutrons, the modified badge contained a 

combination of two TLD chips and two TLND chips for low and intermediate 

energy neutron dosimetry plus an NTA film for fast neutron dosimetry. The MDL 

of this neutron dosimetry is assumed to about the same as that of the NTA film 

alone because most of the neutron dose at Y-12 comes from neutrons above the 

500-keV threshold of the NTA film. In the mid-1980s, the NTA film was removed 
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because of poor film quality, its large MDL, and the labor intensive processing 

requirements (Berger and Lane 1985). (Kerr 2006, pg. 34) 

During casting operations, the decay products of 238U float to the top surface of 

the molten metal and remain as surface residues. These surface residues result in 

an increased exposure potential because of the high beta and photon energies 

associated with the 234Pa nuclide. The 234Pa nuclide emits a number of highenergy photons and has a specific activity that is approximately 2 x 1015 times 

larger than the specific activity of its 238U parent (Henderson 1991). For 234Pa, 

the percentages of photons with energies of 30 – 250 keV and 250 keV or more 

are about 7 and 93%, respectively, and for 238U in equilibrium with its short-lived 

234Th, 234mPa, and 234Pa, the percentages of photons with energies of 30–250 keV 

and 250 keV or more are about 82 and 18%, respectively. Thus, an artificially 

high percentage of photons with energies greater than 250 keV was assumed in 

Table 6.3.4.1-1 for the normal and depleted uranium process areas. This 

produces doses that are claimant favorable because of the increased exposure 

potential to high energy photons from the short-lived 234Pa decay product of 238U. 

(Kerr 2006, pg. 22) 

The neutron dose distributions by energy for the various neutron exposure areas 

at Y-12 are summarized in Table 6.4.2.2-1. By multiplying the recorded neutron 

dose by the area-specific correction factors, the neutron dose equivalent is 

calculated as follows. Consider security personnel who inventory fissile material 

in the Enriched Uranium Storage Area of Building 9212. Assume that the worker 

receives a recorded annual neutron dose of 100 mrem. The corrected neutron 

dose is 151 mrem for neutrons with energies between 0.1-2 MeV, 28 mrem for 

neutron with energies between 2-14 MeV, and 179 mrem for neutrons of all 

energies. These corrections should be applied to both measured neutron dose 

and missed neutron dose. The dose fractions by energy and the associated ICRP 

60 (1990) correction factors for various neutron exposure areas at Y-12 are 

summarized in Table 6.4.2.2-1. (Kerr 2006, pg. 32) 

The dose fractions by neutron energy group and the associated ICRP 60 

correction factors for the various neutron exposure areas at Y-12 are summarized 

in Table F-5. As an example, consider security personnel who inventory fissile 

material in the Enriched Uranium Storage Area of Building 9212 and assume that 

such a person receives a neutron dose of 100 mrem. The corrected neutron dose 

is 151 mrem for neutrons with energies between 0.1–2 MeV and 28 mrem for 

neutrons with energies between 2–14 MeV. Thus, the total corrected neutron 

dose is a total of 179 mrem. These corrections should be applied to both 

recorded dose and missed dose. (Kerr 2006, pg. 45) 

OTIB-0031 (Merwin 2005) provides some additional insight into general assumptions for 

external dose reconstruction. 

Footnotes in Table 1 provide some of these assumptions. 
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c. The TBD indicates that either monthly, quarterly, or annual exchange 

frequencies were used during this period, depending on work locations and the 

potential for exposure. A review of the data indicates that quarterly exchanges 

were predominant; thus, quarterly exchanges have been assumed here to 

calculate the maximum annual missed dose. 

d. The TBD indicates that either quarterly or annual exchange frequencies were 

used during this period, depending on the potential for exposure. A review of the 

data indicates that quarterly exchanges were predominant; thus, quarterly 

exchanges have been assumed here to calculate the maximum annual missed 

dose. (Merwin 2005, pg. 7) 

OTIB-0031 (Merwin 2005) also points out the following. 

As described in ORAUT-OTIB-0020... some cases not having complete 

monitoring data can be processed based on assumptions and methodologies that 

do not involve coworker data. For example, many cases falling in the first 

category above can be processed by assigning ambient external and internal 

doses based on information in the relevant site Technical Basis Documents 

(TBDs). (Merwin 2005, pg. 4) 

As described in the PGDP External Dosimetry TBD...operations at the site began 

in 1952, and in 1953 the site began using dosimeter and processing technical 

support provided by the Oak Ridge National Laboratory (ORNL). Until July 

1960, dosimeters were issued to a limited number of individuals, i.e., those with 

the highest potential for exposure, and the badges were exchanged weekly. After 

that time, dosimeters were assigned to all workers who entered a controlled area, 

and the badges were exchanged and processed on a monthly or quarterly 

schedule. There does not appear to be any significant administrative practice 

that would jeopardize the integrity of the recorded dose of record. (Merwin 2005, 

pg. 4) 

Inconsistencies between Site Profiles 

There is variation among site profiles related to the standard chest thickness and whether an 

adjustment factor is applied. The Paducah medical dose TBD (Turner 2004, pg. 7) states that, 

“For the PA view, an allowance of 5 cm is made for cassette thickness and 26 cm for chest 

thickness between the source and image. Therefore, the SSD is rPA = 183 – 31 = 152 cm. For 

the LAT projection with an assumed chest thickness of 34 cm, rLAT = 183 – 39 = 144 cm.” 

The Fernald medical dose TBD (Chew-2004, pg. 8) states that for units of 25–27 mm, the dose 

increased by a factor of 1.5 and with units of greater than 27 mm, the dose increased by a factor 

of two. 
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The Y-12 TBD (Murray 2006a, pg. 12) states that: “For a given individual, the source-to-skin 

distance (SSD) will be determined largely by the thickness of the patient and the accuracy in 

positioning the patient.” 

The SRS TBD (Scalsky 2005) and INEEL TBD assumes a chest thickness of 26 cm and 34 cm 

for PA chest x-rays and lateral chest x-rays, respectively. OTIB-0006 Revision 2 and Revision 3 

both indicate that the average worker chest size is 22–24 cm (Kathren and Shockley 2005). The 

OTIB recommends that adjustments be applied for larger individuals with chest thicknesses of 

25–27 cm and > 27 cm chest thickness, resulting in an increase in dose by a factor of 1.5 and 2.0, 

respectively. In the absence of site specific data, consistent default assumptions are warranted. 

In the Paducah Environmental Dose TBD (East 2004), there is no mentioned of doing soil 

sampling in order to better understand soil resuspension and its potential impact on inhalation 

dose. Doses from the resuspension of contaminated soil have not been given any attention in the 

Fernald Environmental Dose TBD (Chu 2006), as well. The Y-12 Environmental Dose TBD 

(Murray 2006b, pg. 9) does acknowledge that one pathway for environmental dose is direct 

external radiation from radionuclides in soils and outdoor surfaces, as well as shine from 

buildings and operational units. Surveys were taken at Y-12 as discussed on page 31. 

The second major characterization was performed from 1985 to 1987 and 

involved an outdoor radiological and chemical scoping survey of the 800+ acres 

occupied by the Y-12 facility. This survey included both radiological and 

chemical assessments and included measurements of both gamma ray exposure 

rates and the collection of surface soil samples. (Murray 2006b, pg. 9) 

Dose reconstructors are not given and guidance in the Paducah, Fernald, or Y-12 TBDs to 

quantify the level of exposure associated with contaminated soils. 

There has been no consideration of ingestion dose for Paducah, Fernald, or Y-12 Although 

engineering controls were implemented over time, internal contamination from incidents, reused 

personal protective equipment, and deposition on food and beverages in areas where eating was 

allowed may have led to ingestion. 

The Paducah Occupational External Dose TBD (Kerr 2006) indicates that it was as late as 1998 

before quantitative monitoring of neutron dose was implemented. 

Quantitative monitoring for neutron dose began at PGDP in 1998. TLNDs were 

used in conjunction with appropriate work field calibration factors. Before 1998, 

the beta/photon badge assembly contained a neutron-sensitive element (NTA, 

Eastman Kodak Type 2 film). This element was processed only when requested. 

(NTA film had an energy threshold of about 0.5 MeV.) A review of data does not 

indicate the assignment of neutron dose before 1998. (Turner and Thomas 2006, 

pg. 17) 

At Fernald, the “limitations of NTA film were well documented including an MDL of 

approximately 40 mrem for fast neutrons” (Faust 2004, pg. 18) as noted in the Fernald 
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Occupational External Dose TBD. The Fernald TBD does not mention NTA film in any other 

place than this one site. 

The Y-12 Occupational External Dose TBD (Kerr 2006, pg. 16) indicates that an additional 

ORNL neutron badge containing NTA film and TLND dosimeters was issued to Y-12 personnel 

who were exposed to neutrons. It is noted that: 

An NTA emulsion was included in the same holder used for the Y-12 beta/gamma 

dosimeter until 1980. Between 1980 and 1989, there is a serious gap in the 

neutron dosimetry information for Y-12. 

In general, PGDP, FEMP, Y-12, ORNL, Hanford, SRS and other site profiles discount the use of 

NTA film as an adequate measure of neutron dose. By the late 1960s, ORNL apparently 

abandoned the idea of assigning a thermal neutron dose on the basis of the results of NTA film 

dosimeters. It wasn’t until around 1980–1985 that ORNL implemented the use of TLDs for 

personnel neutron monitoring (Burns and Mohrbacher, pp. 21 and 66). Y-12 became 

increasingly dependent on Y-12, particularly during the years from 1980 to 1989. NTA file was 

still used for measurement of fast neutron dose even after the neutron TLD was implemented. 

(Kerr 2006, pg. 18). The SRS Occupational External Dosimetry TBD (Scalsky 2005) 

distinguishes neutron energies and neutron-to-photon ratios for reactors, fuel fabrication, 

plutonium production, and radionuclide production and calibration (Scalsky 2005). The INEEL 

Occupational External Dose TBD (Rohrig 2004) considers the reactors, the processing plant, 

waste handling operations, calibration sources, and uranium handling. Neutron energy spectra 

and neutron-to-photon ratios for Pu-238 and Pu-239 operations are segregated at SRS (Scalsky 

2005). Further evaluation of neutron-to-photon ratios should include more specific categories 

including neutron sources (RaBe, Cf, etc.), accelerators, early subcriticality experiments, initiator 

development, and neutron spectra from alternate fissile materials. NIOSH/ORAU has proposed 

to include a Pu-238 specific neutron-to-photon ratio in the next revision. 

Exposure geometry is not dealt with in the Paducah Occupational External Dose TBD. There is 

only one mention of exposure geometry. 

The adequacy of dosimetry methods to measure radiation dose accurately is 

determined from radiation type, energy, exposure geometry, and other factors 

described in this section. (Turner and Thomas 2006, pg. 11) 

In the Fernald Occupational External Dose TBD exposure geometry is only mentioned once in 

Table 6-15 on page 24 indicating that the exposure geometry was AP. 

At Y-12, the Y-12 dosimeter system calibrated using A-P laboratory irradiations (Kerr 2006, 

pg. 23. It was reported that the NTA dosimeter response increases with increasing exposure 

angle and TLND response decreases with increasing exposure angle (Kerr 2006, pg. 28). 

At ORNL, the Occupational External Dose TBD (Burns and Mohrbacher 2004, pg. 23) points 

out that 100% AP exposure geometry has been assumed. This is the only mention of exposure 

geometry. The Occupational External Dose TBDs for Y-12 (Murray 2003), SRS (Scalsky 2005), 
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and Hanford (Scalsky 2003) base their default exposure geometry on the compensability or noncompensability of the claim. The MCW (Westbrook 2005) and RFP (Furman & Lopez 2004) 

Occupational External Dose TBDs base default exposure geometries on job titles. The INEEL 

(Rohrig 2004) Occupational External Dose TBDs default to 100% Anterior-Posterior (AP) 

exposure.