September 19, 2005 

Mr. David Staudt 

Center for Disease Control and Prevention 

Acquisition and Assistance Field Branch 

Post Office Box 18070 

626 Cochrans Mill Road – B-140 

Pittsburgh, PA 15236-0295 

Re: Contract No. 200-2004-03805, Task Order 1: Draft Review of the 

NIOSH Site Profile for the Y-12 National Security Complex 

Dear Mr. Staudt: 

S. Cohen & Associates (SC&A) is pleased to submit our draft review of the NIOSH site profile for the 

Y-12 National Security Complex, consisting of our evaluation of “Revision 0” of six technical basis 

documents (TBDs) for site description, and occupational internal, external, medical, and environmental 

dose, respectively. We understand that some or all of these TBDs are under active revision and have 

reflected that understanding with respect to findings and issues that we have identified. 

Given the time required for obtaining DOE clearance of this report for public release, SC&A proceeded to 

brief the Advisory Board on Radiation and Worker Health on the preliminary findings contained in the 

report at the Advisory Board meeting in St. Louis, Missouri, on July 5–7, 2005. The SC&A team also 

held a conference call regarding these findings with NIOSH and ORAU team counterparts. However, a 

more rigorous issue resolution process has not yet been engaged with NIOSH and the ORAU team given 

delays experienced with classification reviews. Given the nature of issues identified in this draft SC&A 

report, we request such a working session as soon as mutually convenient and preferably before the 

October 17, 2005, Advisory Board meeting in Oak Ridge, Tennessee. 

Sincerely, 

John Mauro, PhD, CHP 

Project Manager 

cc: P. Ziemer, PhD, Board Chairperson 

 Advisory Board Members 

 L. Wade, PhD, NIOSH 

 L. Elliott, NIOSH 

 J. Neton, PhD, NIOSH 

 S. Hinnefeld, NIOSH 

 Z. Homoki-Titus, NIOSH 

 A. Brand, NIOSH 

H. Behling, PhD, SC&A 

 J. Lipzstein, PhD, SC&A 

A. Makhijani, PhD, SC&A 

J. Fitzgerald, Saliant 

K. Robertson-DeMers, CHP, Saliant 

S. Ostrow, PhD, SC&A 

K. Behling. SC&A 

 Project File (ANIOS/001/07)
Draft 

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH 

National Institute of Occupational Safety and Health 

Y-12 National Security Complex Site Profile Review 

Contract No. 200-2004-03805 

Task Order No. 1 

SCA-TR-TASK1-0007

Prepared by 

S. Cohen & Associates 

6858 Old Dominion Drive, Suite 301 

McLean, Virginia 22101 

Saliant, Inc.

5579 Catholic Church Road 

Jefferson, Maryland 21755

September 19, 2005 

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.

 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

2 of 203

Document No. 

SCA-TR-TASK1-0007 

Effective Date: 

Draft — September 19, 2005 

S. COHEN & ASSOCIATES:

Technical Support for the Advisory Board on

Radiation & Worker Health Review of 

NIOSH Dose Reconstruction Program Revision No. 0 – DRAFT 

Y-12 NATIONAL SECURITY COMPLEX 

SITE PROFILE REVIEW Page 2 of 203 

Task Manager: 

________________________ Date: ___________ 

Joseph Fitzgerald 

Project Manager: 

________________________ Date: ___________ 

John Mauro 

Supersedes: 

N/A 

 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

3 of 203

TABLE OF CONTENTS 

Acronyms and Abbreviations ..........................................................................................................7 

1.0 Executive Summary...........................................................................................................10 

1.1 Summary of Strengths............................................................................................13 

1.2 Summary of Findings.............................................................................................13 

1.3 Opportunities for Improvement .............................................................................17 

2.0 Scope and Introduction ......................................................................................................19 

2.1 Review Scope.........................................................................................................20 

2.2 Review Approach...................................................................................................22 

2.3 Report Organization...............................................................................................24 

3.0 Assessment Criteria and Methods......................................................................................25 

3.1 Objectives ..............................................................................................................25 

3.1.1 Objective 1: Completeness of Data Sources.............................................25 

3.1.2 Objective 2: Technical Accuracy..............................................................25 

3.1.3 Objective 3: Adequacy of Data.................................................................26 

3.1.4 Objective 4: Consistency Among Site Profiles.........................................26 

3.1.5 Objective 5: Regulatory Compliance........................................................26 

4.0 Site Profile Strengths.........................................................................................................31 

4.1 Completeness of Data ............................................................................................31 

4.2 Adequacy of Data ..................................................................................................31 

4.3 Technical Accuracy/Claimant Favorability ...........................................................32 

4.4 Consistency Among Site Profiles..........................................................................34 

4.5 Regulatory Compliance .........................................................................................34 

5.0 Vertical Issues....................................................................................................................36 

5.1 Issue 1: Unmonitored Support Services and Crafts Workers ...............................36 

5.1.1 Potential Exposure of Y-12 Support Services Workers.............................37 

5.2 Issue 2: Pre-1961 External Dose...........................................................................38 

5.2.1 General.......................................................................................................38 

5.2.2 Tennessee Eastman Corporation................................................................39 

5.2.3 Monitoring from 1948–1949......................................................................40 

5.2.4 Monitoring from 1950–1961......................................................................41 

5.2.5 Missed Photon Dose ..................................................................................42 

5.2.6 Statistical Application for the Assignment of Missed Dose ......................45 

5.3 Issue 3: Radiobioassay and Internal Dose Calculations .......................................48 

5.3.1 General Information...................................................................................49 

5.3.2 Particle Size ...............................................................................................50 

5.3.3 Uncertainties and Detection Limits of the Bioassay Techniques ..............51 

5.3.4 Interferences...............................................................................................54 

5.3.5 Potential for Ingestion Exposure Pathways ...............................................55 

5.3.6 Co-worker Dose Assignment.....................................................................57 

5.4 Issue 4: Other Radionuclides................................................................................64 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

4 of 203

5.4.1 Recycled Uranium .....................................................................................65 

5.4.2 Uranium-233 Production ...........................................................................69 

5.4.3 Thorium Workers.......................................................................................70 

5.4.4 Technetium and Tritium Exposure ............................................................72 

5.4.5 Neptunium..................................................................................................73 

5.4.6 Plutonium and Other Radionuclides..........................................................73 

5.5 Issue 5: Neutron Dosimetry and Exposures..........................................................74 

5.5.1 Neutron Energy Spectra Characterization .................................................75 

5.5.2 NTA Film Response ..................................................................................76 

5.6 Issue 6: Radiation Generating Devices (RGDs) ...................................................78 

5.6.1 X-ray Machines..........................................................................................79 

5.6.2 Accelerators ...............................................................................................80 

5.6.3 Critical Assemblies....................................................................................82 

5.7 Issue 7: Description of Historic Y-12 Operations and Radiological Control 

Practices.................................................................................................................82 

5.7.1 Operational Process Descriptions..............................................................83 

5.7.2 Radiological Controls................................................................................85 

5.8 Issue 8: Occupational Environmental Dose..........................................................89 

5.8.1 Additional Exposure Pathways..................................................................90 

5.8.2 Inhalation of Uranium................................................................................90 

5.8.3 External Exposure......................................................................................96 

5.9 Issue 9: Frequent “Incidental” Sources of Chronic Workplace Radiation 

Exposure ................................................................................................................97 

5.10 Issue 10: Skin and Extremity Doses .....................................................................99 

5.10.1 Acute Beta Doses and Routine Nonpenetrating Doses............................100 

5.10.2 Extremity Exposure .................................................................................101 

6.0 Overall Adequacy of the Site Profile as a Basis for Dose Reconstruction ......................102 

6.1 Satisfying the Five Objectives.............................................................................102 

6.1.1 Objective 1: Completeness of Data Sources...........................................102 

6.1.2 Objective 2: Technical Accuracy............................................................104 

6.1.3 Objective 3: Adequacy of Data...............................................................105 

6.2 Objective 4: Consistency Among Site Profiles...................................................105 

6.2.1 Objective 5: Regulatory Compliance......................................................106 

6.3 Usability of Site Profile for Intended Purpose.....................................................106 

6.3.1 Ambiguous Dose Reconstruction Direction ............................................107 

6.3.2 Inconsistencies and Editorial Errors in the Site Profiles..........................108 

6.4 Unresolved Policy or Generic Technical Issues ..................................................108 

7.0 References........................................................................................................................110 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

5 of 203

LIST OF TABLES

Table 1: One-Year 234

U Committed Equivalent Doses per Bq 234U Present in Urine after 

Weekend Urine Sample, Collected at the End of 1-Year Exposure ........................... 55 

Table 2: One-Year 234

U Committed Equivalent Doses per Bq 234U Present in Urine after 

Weekend Urine Sample, Collected at the End of 1-Year Exposure ........................... 59 

Table 3: Estimated Bone Surface Doses from Recycled Uranium to Workers at the 

Paducah Gaseous Diffusion (Committed Dose Equivalent – CDE in a Year) 

(DOE 2000c)............................................................................................................... 67 

Table 4: Exposure Analysis in the 86-inch Cyclotron Area for January 1, 1952–July 1, 

1952 (Struxness 1953) ................................................................................................ 81 

Table 5: Calculated P/Q for Three Stations at Y-12 Facility (m-3)........................................... 93 

Table 6: Air Sample Data During and After Uranium Chip Fire.............................................. 98 

Table 7: 1961–1962 Average Skin and Penetrating Dose by Worker Category..................... 101 

LIST OF FIGURES 

Figure 1: ORAUT-RPRT-0032 (Kerr 2005a) Attachment 1, Page 136..................................... 39 

Figure 2: Statistical Measures by Quarter from Table 1 of ORAUT-OTIB-0013 

(m = μ, s = σ).............................................................................................................. 46 

Figure 3: Annual Film Badge Monitoring of Y-12 Workers for Gamma Exposure 

(Watkins 2004)............................................................................................................ 47 

Figure 4: One-Year Committed Dose in Sv per Bq Excreted in After Weekend Urine 

Samples after One Year of Work................................................................................ 56 

Figure 5: One-Year Committed Equivalent Doses..................................................................... 57 

Figure 6: One-Year Committed Equivalent Dose for All Organs.............................................. 60 

Figure 7: One-Year Equivalent Doses for Non-Respiratory Tract Organs. ............................... 61 

Figure 8: One-Year Committed Equivalent Doses from Ingestion of Insoluble 

Compounds................................................................................................................. 62 

Figure 9: Energy Response of a Typical Survey Meter with a 1-inch NaI Crystal 

(Ludlum 2000) ............................................................................................................ 97 

 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

6 of 203

ATTACHMENTS 

Attachment 1: Review of the NIOSH Site Profile for the Y-12 National Security Complex ... 121 

Attachment 2: NIOSH Technical Documents Considered During the Review......................... 132 

Attachment 3: Key Questions for NIOSH/ORAU Regarding Site Profile Documents............. 133 

Attachment 4: Summary of Conference Call on SC&A Questions Provided to NIOSH.......... 144 

Attachment 5: Summary of Environmental Conference Call with NIOSH AND SC&A ......... 167 

Attachment 6: Site Expert Interview Summary [Reserved] ...................................................... 176 

Attachment 7: Consistency Between Y-12, Hanford and SRS Site Profiles............................. 177 

Attachment 8: Uranium Process Flow at the Y-12 Plant (DOE 1985)...................................... 198 

Attachment 9: Editorial Errors in theTBD................................................................................. 199 

Attachment 10: Uranium Separation Process (1943–1947)....................................................... 201 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

7 of 203

ACRONYMS AND ABBREVIATIONS 

Advisory Board Advisory Board on Radiation and Worker Health 

A-P Anterior-Posterior 

AEC Atomic Energy Commission 

AMAD Activity Median Aerodynamic Diameter 

AWE Atomic Weapon Employer 

Bq Becquerel 

Calutron California University Cyclotron 

CCCC Carbide and Carbon Chemicals Corporation 

CCCD Carbide and Carbon Chemicals Division 

CDC Centers for Disease Control and Prevention 

CEF Critical Experiments Facility 

CER Center for Epidemiologic Research 

CFR Code of Federal Regulations

Ci Curie 

DCF Dose Conversion Factor 

DNFSB Defense Nuclear Facility Safety Board 

DOE Department of Energy 

dpm Disintegrations Per Minute 

DU Depleted Uranium

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 

2000 

ET Extrathoracic Region 

EU Enriched Uranium

GSD Geometric Standard Deviation 

HEU Highly Enriched Uranium 

HP Health Physics 

ICRP International Commission on Radiological Protection 

IMBA Integrated Modules for Bioassay Analysis 

IREP Interactive RadioEpidemiological Program

keV Kiloelectron Volt (a unit of energy) 

kVp Kilovolt potential 

 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

8 of 203

LD Detection Limit 

μCi Microcurie 

mA Milliampere 

mAs Milliampere second 

MDA Minimum Detectable Activity 

MDL Minimum Detectable Level 

MeV Million Electron Volts 

μm Micrometers

MMES Martin Marietta Energy System

mR Milliroentgen 

mrad Millirad 

mrem Millirem

NCRP National Council on Radiation Protection and Measurements 

NIOSH National Institute for Occupational Safety and Health 

NTA Eastman Kodak Nuclear Track Film Type A 

NU Natural Uranium

OCAS Office of Compensation Analysis and Support 

ORAU Oak Ridge Associated Universities

ORISE Oak Ridge Institute for Science and Education 

ORNL Oak Ridge National Laboratory 

OTIB Technical Information Bulletin 

PA Posterior-Anterior 

PAL Plant Action Level 

PFG Photofluorography 

PIC Pocket Ionization Chamber 

POC Probability of Causation 

PNL Pacific Northwest Laboratories 

ppm Parts Per Million 

RadCon Radiological Control 

RDGs Radiation Generating Devices 

RU Recycled Uranium

SC&A S. Cohen and Associates 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

9 of 203

SRS Savannah River Site 

TBD Technical Basis Document 

TDEC Tennessee Department of Environment & Conservation 

TEC Tennessee Eastman Corporation 

TIB Technical Information Bulletin 

TLD Thermoluminescent Dosimeter

TLND Thermoluminescent Neutron Dose 

TRU Transuranics 

Y-12 Y-12 Complex National Security Complex 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

10 of 203 

 

1.0 EXECUTIVE SUMMARY 

During its April 25–27, 2005 meeting, the Advisory Board on Radiation and Worker Health 

(Advisory Board) made the decision to expedite the conduct of this review to support its review 

of several Special Exposure Cohort (SEC) petitions submitted by Y-12 National Security 

Complex (Y-12 or Y-12 Plant) workers. This review was conducted between May 3 and 

June 30, 2005, by an S. Cohen and Associates (SC&A) audit, and the initial audit findings were 

presented to the Advisory Board at its meeting of July 5–7, 2005, in St. Louis, Missouri. The St. 

Louis presentation is provided as Attachment 1. 

This review included a classified review at Y-121

of onsite records and interviews of site experts 

by Q-cleared SC&A personnel as well as conference calls between Oak Ridge Associated 

Universities (ORAU) and SC&A counterparts regarding specific Technical Basis Documents 

(TBDs) that make up the Y-12 site profile. 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). As “living” documents, TBDs are constantly being revised as 

new information, experience, or issues arise; for Y-12, this is the case for a particularly critical 

TBD, Y-12 National Security Complex - Occupational Internal Dose, which was being revised 

substantially while this review was underway. In addition to this TBD, other Y-12 technical 

basis documents that were reviewed by SC&A include those for Site Description, Occupational 

Medical Exposure, Occupational Environmental Dose, and Occupational External Dose. A 

complete list of the Y-12 TBDs, as well as supporting Technical Information Bulletins (TIBs), 

that were reviewed by SC&A is provided in Attachment 2. 

The Y-12 Plant is one of the oldest facilities in the nuclear weapons complex, having begun 

operation in 1943 and, by the end of World War II, enriching the uranium ultimately used in the 

Hiroshima weapon. From its original electromagnetic isotope separation process to a 

longstanding mission of producing enriched and depleted uranium components for the nuclear 

weapons stockpile, Y-12 has a long history of handling large quantities of uranium. In its 

diverse operations spread out over some 531 buildings on 811 acres, up to 40,000 workers during 

the war years, and a fraction of that in later years, have been employed in a broad variety of job 

categories, including chemical operators, machinists, pipefitters, assemblers, janitors, 

electroplaters, electricians, maintenance employees, and various crafts and specialty trades. 

They have handled uranium, thorium, neptunium, and other production materials in various 

chemical and metal forms, and in many cases, with residual amounts of contaminants from

recycled feed (e.g., plutonium, technetium, americium) or daughter products (e.g., radium-224 

and radon from arc melting of uranium and thorium). These materials have been variously cast, 

rolled, shaped, and machined by thousands of workers over five decades using radiation

protection practices that evolved intermittently over that timeframe. 

1 While the review was conducted in a secure area at Y-12, many of the documents reviewed were

unclassified or designated as “Unclassified Controlled Nuclear Information “ or “Official Use Only.” All material 

generated from worker interviews and document reviews were submitted to classification screening to assure that 

no sensitive information was inadvertently included in this report. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

11 of 203 

This review found that the NIOSH site profile (and its constitutent TBDs) for Y-12 to be an 

adequate accounting of most, but not all, of the uranium exposure and dosimetric history of the 

plant, falling short in fully characterizing several key underlying issues fundamental to guiding 

dose reconstruction. 

For example, chronic uranium uptake is acknowledged for all workers at Y-12, but the 

implications of this chronic radiation exposure are not addressed in terms of the “missed dose” it 

may represent. Likewise, the uncertainties in the bioassay techniques and detection limits used 

to quantify internal dose are significant issues in dose reconstruction that are not fully addressed 

in the TBD. A number of issues related to the intake parameters applied in the TBD need to be 

addressed. For example, the NIOSH intake model does not treat the uncertainties inherent in 

bioassay measurements to detect intakes from urine samples after exposure to Type S uranium,

as found in “high-fired” oxides. The intake model does not take into consideration the 48-hour 

delay imposed on obtaining urine samples following chronic exposures, which can result in an 

under-estimation of the internal dose by a factor of about 2–4, depending on lung clearance 

absorption types. In addition, the intake model does not consider the ingestion pathway despite 

the fact that eating, drinking, and smoking were routine in contaminated operating areas of the 

plant until 1989. With the ongoing revision of the Internal Dose TBD and other ongoing 

assessments, it is understood that some of these issues may already be under review by NIOSH. 

The guiding assumption for assigning pre-1961 external dose for unmonitored workers―that the 

relatively few co-workers that were badged before 1961 represent the “maximally exposed 

individuals”―has not been corroborated by NIOSH. The TBD accepts that co-worker doses can 

be used as a surrogate for unmonitored workers, even though only a relatively small fraction 

(2%−23%) of workers were badged during that period and because it was the stated facility 

policy to badge those workers whose jobs could expose them to radiation exposure at or in 

excess to 10% of the radiation protection guides. However, this claim appears to be 

uncorroborated in the TBD with no referenced badging procedures, personnel selection records, 

or job-specific exposure survey information. The fact that line supervisors, not health physics 

personnel, made badging decisions raises doubts regarding this guiding assumption in the 

absence of such substantiation. 

Traditional uranium production exposures (i.e., depleted uranium, enriched uranium, and highly 

enriched uranium) are characterized, but potentially more hazardous thorium, 233U, and 

transuranic exposures are not. Few distinctions are made in the TBDs between worker job 

categories despite these distinctions being critical. Chemical operators and machinists received 

relatively frequent bioassays and lung counts, while maintenance employees, janitors and 

security guards did not (although the latter jobs entailed access to all operational areas and 

exposure to radioactive contaminants). These job categories likely constitute a significant source 

of potential “missed dose” given the lack of personnel monitoring, job assignment records, and 

accounts from these workers that their routine work involved access to all onsite operational 

areas and may have involved activities leading to high exposure potential. 

No mention is made of high-fired (black oxide) uranium workers who were on urinalysis as their 

primary bioassay, with significant intakes not detected by the techniques that were used at 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

12 of 203 

different times, until the late-1990s, when it was discovered that the highly insoluble nature of 

this material actually required routine fecal sampling as well. Likewise, uranium (and thorium)

chip fires are not addressed although they were frequent and filled the operating area with metal 

fumes. 

The importance of accounting for potential “missed” dose from Y-12’s history of prevalent 

workplace contamination is not trivial. Like other uranium production facilities of that era in the 

Oak Ridge complex – i.e., Fernald, Portsmouth, Paducah, and Oak Ridge Gaseous Diffusion 

Plant (K-25) – uranium was not seen as a radiological hazard of significance from the 1940s 

through the 1980s. Widespread contamination was persistent, radiological controls were few, 

and chronic low levels of uranium uptake were assumed. Eating, drinking, and smoking in 

contaminated uranium processing areas at Y-12 were not prohibited until 1988–1989. It was 

standard practice until that time to wait over a weekend to provide monthly urinalysis samples to 

permit the more soluble uranium to be “flushed out” before counting. Airborne and surface 

contamination levels for uranium repeatedly exceeded Y-12’s “Plant Action Levels” by orders of 

magnitude without effective corrective action. Respirators were “recommended” for certain 

jobs, but were not fit-tested and were worn at worker’s (and supervisory) discretion. The 

persistence of these conditions, and lack of compliance with Department of Energy (DOE) 

requirements and its longstanding “as-low-as-reasonably-achievable” mandate, led to a formal 

complaint by the Defense Nuclear Facilities Safety Board (DNFSB) to the Secretary of Energy in 

1993. The plant suspended operations in 1994 for 4 years for procedural compliance issues, 

among other reasons. A full characterization in the TBD of such workplace radiological 

conditions and their implications for past dosimetry and current dose assessment is lacking. 

It is in this context that SC&A considered it important to corroborate the “written record” and 

stated practices with a review of both historic documentation, as well as interviews with 

longstanding workers. As noted above, what was formally prescribed by either DOE or Y-12 

policies and requirements was not necessarily implemented at the plant in practice. Despite 

substantial findings of noncompliance with radiation protection requirements and procedures, 

Y-12 did not fully implement changes in radiological practices until the operational standdown 

of 1994. The degree to which NIOSH confirmed that the improvements in written radiation 

protection practices, as described in the TBD, were in fact implemented is not documented in the 

TBD or in NIOSH records of worker interviews. 

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 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

13 of 203 

review criteria. Several of the issues were designated as Findings, because they represent 

deficiencies in the TBDs that need to be corrected, and which have the potential to substantially 

impact at least some dose reconstructions. 

1.1 SUMMARY OF STRENGTHS 

NIOSH has supplemented the Y-12 site profile documents with a number of TIBs that provide 

further direction to the dose reconstructor. These six TIBs, listed in Attachment 2, all have direct 

bearing on the Y-12 dose reconstruction process. SC&A has reviewed these carefully and has 

found each to be of significance in providing help to the dose reconstructor. These documents 

were beneficial in understanding the application of the six Y-12 TBDs to the dose reconstruction 

process. 

NIOSH has provided helpful information on solubilities of uranium compounds and their 

absorption types, as well as a thorough analysis of the in-vitro bioassay methods for uranium and 

other radionuclides that were used during the different eras of Y-12 operations. NIOSH has also 

included a good description of the chest counting methods used for uranium and thorium 

monitoring. 

NIOSH has considered the needed adjustments that had to be made to the dose reconstruction 

procedures as changes occurred at Y-12 in dosimetry technology, response functions, and 

changes in radiation fields associated with major work locations. NIOSH has addressed the 

gamma and neutron energy spectrums and the associated detector responses in order to refine 

dose potential in the varied Y-12 operations. At the time of the preparation of this report, ORAU 

had conducted recent visits to the Y-12 Plant to collect information which should address some

of the gaps in the internal dosimetry TBD identified in this report. 

NIOSH has developed Technical Information Bulletin: Dose Reconstruction from 

Occupationally Related Diagnostic X-ray Procedures (Kathren 2003) to augment the 

information regarding x-ray unit output and associated dose. Although this was not made part of 

the occupational medical TBD, Rev. 01 (Murray 2004b), it is of valuable assistance to the dose 

reconstructor. 

1.2 SUMMARY OF FINDINGS 

Finding 1: External monitoring of Y-12 workers was limited to 2%−23% of the workers during 

the period from 1950 to1961, and no external dosimetry data are readily available for workers for 

the period from 1948 to1950. Based on worker interviews, routine bioassay sampling among 

support workers who had access to multiple areas onsite was not implemented prior to 1994. 

The implications of these limitations in the external dosimetry and bioassay programs may be 

significant in terms of the potential for missed dose, given the large numbers of workers 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

14 of 203 

 

involved,2 and the radiation sources to which they may have been exposed. Support services 

workers (e.g., janitors, security, maintenance personnel, and crafts such as pipefitters, 

boilermakers, and welders) were routinely exposed to high-enriched and depleted uranium,

insoluble (“high-fired”) uranium oxides, thorium, and transuranic contaminants, as well as 

various source terms associated with Oak Ridge National Laboratory’s (ORNL) research and 

isotope production operations co-located at Y-12. Assigned to cleaning up contaminated plant 

areas, servicing machining equipment and lathes, and fixing piping and pressure vessels, these 

workers had free access to all operational areas. Workers indicate that they were given daily 

work assignments in different plant locations, for which records were apparently not maintained. 

While the brevity of this review did not permit an extensive records search, NIOSH needs to 

verify to what extent any records exist for the historic exposure and work assignments of support 

services and crafts workers. In light of these findings, the TBDs should be especially attentive to 

how dose reconstruction will be performed for support services and crafts workers who were not 

routinely badged before 1961, or bioassayed before 1994. A similar issue exists for 

decontamination and decommissioning workers, as well as construction workers at the plant 

following the initiation of plant operations. 

Finding 2: The guiding assumption for assigning pre-1961 external dose for unmonitored 

workers – that the relatively few co-workers that were badged represented the “maximally 

exposed individuals” – has not been corroborated by NIOSH. There were several factors 

associated with the Radiological Control (RadCon) program, which would influence which 

workers were badged and how the results were recorded. Line supervisors made all badging 

decisions for groups and individuals within groups (Patterson 1957) in an era where production 

often took precedence over safety, which raises doubts over how much weight should be 

afforded these management assignments. Historic discrepancies in recording badge readings 

also raise questions. For example: (1) Beginning in mid-1956, all film badges reading “below 

the minimum detectable [were directed to be] recorded as the average of the minimum detectable 

reading and zero, instead of being recorded as the minimum detectable” (West 1956a); and 

(2) “Some fraction” of overall Y-12 film badge results were found to be flawed during 1950–

1955 due to incorrect assignment of badges based on an incorrect determination of whether 

gamma or beta radiation exposure was predominant for a worker (West 1991). These and other 

issues need to be evaluated to verify that the “maximally exposed individuals” actually received 

monitoring. Finally, NIOSH used statistical information on exposure of population groups to 

estimate doses to unmonitored individuals in a way that is not necessarily claimant favorable and 

does not make provisions for considering supplementary information (e.g., from interviews) in 

adjusting the statistical parameters used for inferring missing doses. SC&A is concerned that 

there is not an adequate explanation or reason to assume that the exposures experienced by a 

monitored worker population group can be used as a surrogate for unmonitored workers. 

Finding 3: The internal dose TBD does not adequately address the substantial potential for 

missed dose, due to historic bioassay practice at Y-12, and applies an incomplete and flawed 

2

 SC&A’s records review did not identify any documentation that categorized these workers by job 

category or location of work within the plant; however, some information may be available in personnel 

employment records. In addition, worker interviews indicated that support and crafts workers represented a 

significant proportion of Y-12 workers. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

15 of 203 

intake model. It does not sufficiently assess the dosimetric implications of Y-12 workers being 

continuously exposed to low-level chronic uranium exposures and having this exposure 

effectively treated as operational “background” radiation for health physics purposes. The 

ingestion exposure pathway is not considered, despite it being a likely route of exposure. Coworker dose assignments do not consider the shortcomings of the uranium urinalysis database, 

variability in material type, bioassay technique uncertainties, and sampling methodology. 

NIOSH’s use of a 50th percentile intake rate to calculate doses for unmonitored workers based on 

a lognormal distribution of co-worker urinalysis data for 1950–1988 is not considered claimantfavorable, given the measured high urine activity concentrations. Likewise, the uncertainties in 

the bioassay techniques and detection limits used to quantify internal dose are significant issues 

in dose reconstruction and are not fully addressed in the TBD. For example, the NIOSH intake 

model does not treat the uncertainties inherent in bioassay measurements to detect intakes from

urine samples after exposure to Type S uranium, as found in “high-fired” oxides. It also does not 

address the implications of a 48-hour absence from the workplace by the worker following a 

weekend before routine urine samples were taken; without an appropriate adjustment, this would 

lead to underestimations by the TBD model of a factor of 2–4 as a function of lung clearance 

type. Measured particle size values should be used over a default value of 5 μm (as called for by 

42 CFR Part 82), particularly at Y-12, where particle size has been found to range from 

1–10 μm.

Finding 4: The internal dose TBD is incomplete in its review of the historic dose contribution 

of radioisotopes other than uranium. These radionuclides include 3

H, 90Sr, 99Tc, 210Po, 228Th, 

232Th, 239

Pu, 241Pu, 237Np, 233U, and 241Am. Some of these radionuclides were associated with 

research and development activities, while others were handled in production, either as a source 

material or as a contaminant, e.g., from recycled uranium. Given their potential radiation 

exposure significance, thorium, 233U, and transuranic handling are of particular importance to the 

TBD. While some of these operations are associated with ORNL missions at Y-12, a number of 

Y-12 workers (e.g., janitors, maintenance, and crafts personnel, as noted above) supported these 

operations and were potentially exposed to these sources. NIOSH has recognized this issue and 

ORAU has an ongoing effort to better characterize the radiological significance and exposure 

records associated with thorium, 233U, and recycled uranium. 

Finding 5: SC&A has reservations about the assumptions made regarding the ability of nuclear 

track Type A emulsion (NTA) film to adequately characterize Y-12 worker exposure due to the 

limited amount of neutron spectral measurements available. Neutron radiation sources are not 

adequately defined in the TBD for all potential neutron exposure conditions at Y-12, including 

spontaneous fission neutrons, moderated (alpha, neutron) sources in solutions/compounds, 

subcritical and critical assemblies, and moderated neutrons from the 86-inch cyclotron. The 

assumption that most neutron exposure at Y-12 was from neutron energies greater than 500 keV, 

and thus would be detected by the NTA film, is not adequately substantiated. In fact, Griffin et 

al. (1979) and Wilson et al. (1990) state that NTA is a poor detector of film energies between 

500 keV and 800 keV. In general, the NTA film response decreases logarithmically as the 

energy decreases from 1,000 keV to 500 keV. Therefore, NTA may seriously underestimate the 

true neutron dose, and appropriate correction factors for this underestimate should be applied. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

16 of 203 

 

Finding 6: The TBD intermittently mentions the use of various radiation generating devices at 

the Y-12 plant;3

 however, it does not take into account all radiological hazards associated with 

operations of these units. There is no discussion of how film badge response to low-energy 

x-rays may impact the recorded dose or the correction factors that are necessary to compensate 

for low-energy photon exposures from x-ray fields. Although the TBD recognizes that the 86-

inch cyclotron was used for isotope development, there is no discussion addressing the internal 

dose associated with the isotopes generated from this operation. Finally, other accelerators, 

although mentioned in technical reports and ORAUT-RPRT-0033 (Kerr 2005b), are not included 

in the TBD. Although some of these devices may have been operated by ORNL personnel at the 

Y-12 complex, there is still a potential for exposures to Y-12 personnel, which should be 

included in the TBD. 

Finding 7: The TBDs do not go into enough depth on the varying and changing nature of Y-12 

operations and work environments to provide the dose reconstructors with the specific 

knowledge that is needed for specific group or individual dose reconstructions. It is important to 

adequately understand the potential hazards and dose potential for each operation performed. 

Changes in operations that occurred over varying periods and the changing mission of facilities 

and buildings can also have impacted the doses that individual workers or groups of workers 

might have received. The ability to reconstruct worker doses in these many unique and varying 

types of functions is dependent on a better-developed description of individual or group activities 

and how their operations were modified by redirection in the missions of the facilities.

Finding 8: The occupational environmental dose TBD (Ijaz and Adler 2004) leaves two 

important exposure pathways inadequately addressed. Inadvertent ingestion of radioactively 

contaminated soil and inhalation of radionuclides other than uranium need to be more fully 

developed, if adequate dose reconstructions from these sources are to be effectively taken into 

account. The ORNL facility handled irradiated thorium and radioactive lanthanum and 

plutonium, and was near enough to Y-12 to cause exposure to outdoor Y-12 workers from its 

airborne emissions. The TBD does not address worker exposures to these potential sources. In 

addition, no data are furnished regarding the activity median aerodynamic diameter (AMAD) or 

the chemical form of airborne uranium. Further, the empirical approach used to derive 

atmospheric dispersion factors for the purpose of reconstructing outdoor exposures of 

unmonitored workers is based on several premises that are not explicitly stated and need to be 

explored and justified before this approach can be accepted. Specifically, SC&A questions the 

decision to reconstruct outdoor exposures to airborne uranium using the atmospheric dispersion 

factors as derived in the TBD. We also question the validity of deriving outdoor inhalation 

exposures to uranium using source terms and empirically derived atmospheric dispersion factors 

when actual airborne uranium concentrations are available at outdoor receptor locations. 

Finding 9: The TBDs need to better characterize frequent “incidental” sources of workplace 

radiation exposure. For example, uranium chip fires, which were relatively frequent events 

(sometimes several times a shift, according to workers interviewed), exposed everyone in the 

3

 Including the 86-inch cyclotron, the 9201 proton-accelerating cyclotron, secondary ion mass 

spectrometers, x-ray photoelectron spectrometer, enclosed beam diffraction equipment, and various x-ray devices;

with energies ranging from 15 keV to 9 MeV. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

17 of 203 

work area to an acute intake of uranium fumes. In other recorded instances, exhaust fans were 

inoperative, causing a back-flow of contaminated air through ductwork. Radium and its progeny 

were preferentially vaporized during molten uranium and thorium casting operations (due to 

having a lower melting point), leading to sharply elevated airborne radioactivity levels (Ammann 

1960; West 1965). It is not clear whether the frequency and sensitivity of the historic Y-12 

bioassay program would have detected these “spike” releases, particularly where urinalysis was 

delayed or unduly relied upon for Type S uranium oxides. Such releases, in particular, would 

need to be addressed where routine air sampling data is used to estimate potential worker 

uptakes. 

Finding 10: Extremity and skin doses should be given attention in the external dose TBD. 

Depleted uranium metal handling has associated high beta radiation fields leading to elevated 

hand and arm exposures of material handlers. Similarly, 232Th has beta-emitting progeny that 

present a radiological hazard for direct handling. Based on a sampling of quarterly health 

physics reports from building 9206 in the early 1960s, the average quarterly skin dose ranged 

from 100–400 mrem (0.4–1.6 rem/year) for three categories of monitored workers (mechanical 

operators, product certification, and chemical operators) in that building. Since only a few 

workers wore ring dosimeters, it is important to more fully characterize potential beta/gamma

fields and worker handling practices and geometries. 

1.3 OPPORTUNITIES FOR IMPROVEMENT 

Support Workers: Support service workers were often assigned to multiple areas of the site, 

resulting in potential exposure to any number of processes and radionuclides. Consideration 

should be given to the adequacy of internal and external monitoring, and availability of reliable 

dose and job assignment records, for support personnel with respect to the potential exposure 

conditions. 

Incomplete Personal Monitoring Data: The TBD should further substantiate the assumptions 

used to assign external dose to workers who were not monitored prior to 1961. Consideration 

should be given to Radiological Control (RadCon) practices, data discrepancies, and recording 

practices. The applicability of worker population data to individual unmonitored workers should

include claimant-favorable assumptions and make provisions for considering supplementary 

information. 

Other Radionuclides: The dose contribution from 99Tc, 232Th and daughters, 233U and 

daughters, transuranics, and tritium as an impurity in recycled uranium and/or process materials 

should be elevated in terms the relative impact on internal, external, and environmental dose 

reconstruction. 

Neutron Monitoring: NIOSH should validate the assumption that all sources of neutron 

exposure, including moderated neutron sources, are of sufficient energy to be detected by the 

NTA film. Correction factors should be applied where appropriate based on the analysis. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

18 of 203 

Radiation Generating Devices: NIOSH should include a discussion on radiation generating 

devices (RGDs) used at the Y-12 Plant encompassing the range of devices, their energy 

potential, and exposure histories, where available. 

Environmental Dose: NIOSH should include inadvertent ingestion from radioactively 

contaminated soil and exposure to isotopes other than uranium in the assessment of 

environmental dose. The use of empirically determined atmospheric dispersion factors for 

estimating the outdoor doses of unmonitored workers should be based on explicitly stated and 

justified assumptions. 

Incidental Exposures: NIOSH should create a list of high-risk jobs and incidents for 

consideration as a complement to the site profile to inform the dose reconstructor during the 

individual dose reconstruction process. Such a list would be particularly pertinent to incidental 

exposures, such as chip fires, which occurred frequently and may have resulted in exposures not 

addressed in the TBD. 

Shallow and Extremity Dose: NIOSH should develop a methodology for assessing extremity 

and skin dose to personnel in the immediate vicinity of uranium, thorium, or other radioactive 

material. 

Dose Calculation Example: A text example of a hypothetical individual dose reconstruction 

using recorded records, missed dose assignment, and dose assignments when dosimeters read 

zero dose should be provided in the TBD. 

Use of Site Expert Input: NIOSH should make a greater effort to take into account site expert 

information and investigate worker accounts. The on-site, first-hand experience of site experts 

enables them to provide original perspectives and information concerning site practices and 

exposure histories. NIOSH has incorporated a limited amount of worker input into the latest 

versions of the TBD. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

19 of 203 

2.0 SCOPE AND INTRODUCTION 

During its April 25–27, 2005 meeting, the Advisory Board made the decision to expedite the 

Y-12 site profile review. The urgency for SC&A to conduct this review was largely influenced 

by several Special Exposure Cohort (SEC) petitions that were under review by the Advisory 

Board at the time, as well as by NIOSH. The review consisted of both classified and unclassified 

information pertaining to the Y-12 site, with the unclassified review beginning on April 27, 

2005, and the classified review and site expert interviews occurring from June 2–15, 2005. 

The Y-12 Plant (currently the Y-12 National Security Complex) poses some unique challenges 

given its lengthy operating history (over 60 years), and classified weapons material and 

components fabrication history. The specific mission of Y-12 was to separate fissionable 

isotopes of uranium for use in atomic weapons. However, during its long history, it has 

processed 233

U, neptunium, and thorium, and has had plutonium, tritium, and other radionuclides 

in its operations either in production, research, or as a contaminant. The plant has had a 

succession of missions before, during, and after the Cold War, including electromagnetic 

separation of uranium (1943–1947), production of weapon secondaries, refurbishment and 

dismantling of nuclear weapon components, storage and processing of uranium and lithium, and 

production support to the weapons laboratories. From 1943–1947, the Y-12 Plant was operated 

by the Tennessee Eastman Corporation. Unlike many other DOE and DOE-predecessor 

facilities, Y-12 has had only a limited number of different management and operating 

contractors, including the Tennessee Eastman Corporation from 1943 to 1947, and Union 

Carbide, Martin Marietta (becoming Lockheed Martin), and BWXT, successively, since 1947. 

Radiation protection monitoring evolved at the plant as DOE’s requirements and safety 

management practices (and those of its predecessor agencies), and prevailing national radiation 

protection standards and practices, themselves, evolved over time. Given the secrecy 

surrounding its activities and DOE’s longstanding self-regulation under the Atomic Energy Act 

of 1954, radiation protection practices at Y-12 did not always keep pace with accepted practices 

in the regulated commercial sector or even at other DOE facilities. In particular, this was the 

case for uranium operations, its core mission activity. In the early years at Y-12 (to the 1980s), 

uranium was not viewed as a radiation hazard of great significance, particularly in contrast with 

plutonium, thorium, and other radionuclides. As noted in the Internal Dose TBD, early health 

physics practices recognized the “nephrotoxicity” of uranium to the kidneys as the limiting 

hazard for low-enriched and depleted uranium. Recognition that extensive direct handling of 

uranium metal and solutions did lead to cumulative doses of concern, led Y-12 to begin badging 

certain workers in higher exposure jobs, such as chemical and mechanical operators. Similar 

recognition of potential intakes of airborne uranium led to urinalysis being provided for a select 

group of workers beginning in 1954. However, these numbers grew slowly until the 1958 

criticality and other circumstances led to a broadening of personnel monitoring beginning with 

plant-wide badging in 1961. 

However, despite improved monitoring and a health physics program seemingly responsive to 

new or unusual exposure situations, the Y-12 radiation protection program fell short for 

controlling uranium exposure. Eating, drinking, and smoking were permitted in the uranium
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

20 of 203 

production areas until 1988–1989. Radiation contamination egress monitoring was not instituted 

until the early 1990s. In-plant radioactive contamination zoning was inadequate, and any worker 

could enter most process areas without restriction. Respirators were available to workers, but 

discretionary for all but a few high-exposure jobs; often it was the supervisors who decided 

whether they would be worn. Plant contamination action levels were exceeded frequently, 

sometimes by orders of magnitude, without attention to root causes and long-lasting remedies. 

Operating program unresponsiveness to identified audit deficiencies led both DOE and the 

DNFSB to cite Y-12 practices in 1988 (DOE 1988) and 1993 (DNFSB 1993), respectively. 

These complaints, coupled with the 4-year standdown ending in 1998, finally brought about 

fundamental changes in uranium practices and the full institution of an ALARA philosophy to 

operations. 

The implications of this history are relevant to the TBDs, although not addressed in any 

meaningful manner. For example, chronic levels of radiation exposure from uranium handling, 

while recognized, were not seen as a concern and, therefore, not given much attention in the 

Y-12 dosimetry program. Therefore, missed dose due to chronic uranium exposure should be 

one focal point of the TBDs. Supervisors were typically given latitude to make decisions 

regarding which of their workers were provided badging, bioassay, and respiratory protection, 

and how radiation jobs were performed. Therefore, NIOSH needs to be cautious with respect to 

the assumptions employed in dose reconstructions regarding radiation protection practices 

employed at Y-12. With the production exigencies of the Cold War, production was often the 

first priority and workers were kept at their contaminated workstations for almost the entire 

workday. Therefore, attention must be given to exposure pathways such as ingestion. These and 

other operational considerations were considered by SC&A in determining radiation exposure 

issues that would be influential in ascertaining possible gaps and inadequacies in Y-12’s historic 

radiation dosimetry practices and dose records. In turn, such considerations were the basis of 

evaluating the technical adequacy and completeness of the NIOSH TBDs. 

2.1 REVIEW SCOPE 

Under the Energy Employees Occupational Illness Compensation Program Act of 2000 

(EEOICPA) and Federal regulations defined in Title 42, Part 82, Methods for Radiation Dose 

Reconstruction Under the Energy Employees Occupational Illness Compensation Program, of 

the Code of Federal Regulations (42 CFR Part 82), the Advisory Board on Radiation and Worker 

Health is mandated to conduct an independent review of the methods and procedures used by the 

NIOSH and its contractors for dose reconstruction. As a contractor to the Advisory Board, 

S. Cohen and Associates (SC&A) has been charged under Task 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. 

This report provides a review of the following six documents related to historical occupational 

exposures at the Y-12 Site: 

• ORAUT-TKBS-0014-1, Technical Basis Document for the Y-12 National Security 

Complex – Y-12 Site Profile (Murray 2004a) 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

21 of 203 

• ORAUT-TKBS-0014-2, Technical Basis Document for the Y-12 National Security 

Complex – Site Description (Jessen 2005) 

• ORAUT-TKBS-0014-3, Technical Basis Document for the Y-12 National Security 

Complex – Occupational Medical Dose (Murray 2004b) 

• ORAUT-TKBS-0014-4, Technical Basis Document for the Y-12 National Security 

Complex – Occupational Environmental Dose (Ijaz and Adler 2004) 

• ORAUT-TKBS-0014-5, Technical Basis Document for the Y-12 Site –Occupational 

Internal Dose (Rich and Chew 2005) 

• ORAUT-TKBS-0014-6, Technical Basis Document for the Y-12 Site –Occupational 

External Dosimetry (Kerr 2003) 

These documents are supplemented by technical information bulletins, which provide additional 

guidance to the dose reconstructor. A complete list of these documents is available in 

Attachment 2. 

SC&A, in support of the Advisory Board, has critically evaluated the Y-12 Site Technical Basis 

Documents (TBDs) in order to: 

• 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 

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

review was conducted in accordance with 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 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

22 of 203 

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 used to determine 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, co-worker data and workplace 

monitoring data, and process description information or source term data. 

NIOSH/ORAU has acknowledged that ORAUT-TKBS-0014-5, Revision 01 (Rich and Chew 

2005) needs to be revised in order to encompass newly retrieved documentation on thorium,

recycled uranium, and in-vivo counting (see Attachment 4). NIOSH has stated that they are in 

the process of modifying the TBD in order to incorporate additional information and to further 

explain the technical basis for their assumptions. A date of release for Revision 2 to the Internal 

Dose TBD had not been announced at the time of preparation of this report. To our knowledge, 

there are no immediate plans to update other portions of the Y-12 site profile (albeit, verbal 

comments have been made regarding some likelihood that the current TBDs will be 

supplemented with treatment of extremity exposures). 

2.2 REVIEW APPROACH 

In past reviews of other TBDs, SC&A’s approach has been to engage a team of scientists where 

every member independently assessed the entire document before coming together for a critical 

open discussion. Due to the classified nature of many of the documents supporting the TBDs, 

this approach was not possible. Instead, the review of the TBDs proceeded down two paths, one 

unclassified and the other classified. To accommodate the need for review of classified 

documents, two members of the SC&A team with Q-clearances inspected classified and sensitive 

documents and conducted classified site expert interviews. They prepared an analysis that was 

not reviewed by the rest of the team, but was submitted for declassification. At the time of the 

preparation of this report, the classified report was not entirely declassified, and therefore some

of the information contained therein is not incorporated into this report. However, we believe 

that this report is substantially complete. 

The classified review itself, while critical to a broader understanding of key dose assessment 

models and assumptions, was constrained in a number of important ways. First, due to time 

limitations and the logistics of scheduling appropriately cleared SC&A members, the review was 

delayed until June 2, 2005, and site expert interviews were not completed until June 15, 2005. 

Scheduling involved two separate organizations at Y-12 to allow for interview of safety, 

production, and security personnel. Second, notes taken during the site expert interviews and 

classified document reviews were only partially cleared and provided to SC&A in time for their 

use as a reference during the development of this report. Third, normal discourse was precluded 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

23 of 203 

between SC&A team members evaluating the same issues on both sides of the classification 

divide. As a result, we were unable to collectively discuss and integrate our findings. 

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

completeness of data collected by NIOSH, the adequacy of existing Y-12 personnel and 

environmental monitoring data, and the evaluation of key dose reconstruction assumptions. Site 

expert interviews were conducted in a secure facility at Y-12, and Y-12 derivative classifiers 

screened all notes taken. Records reviews were conducted at the Y-12 classified and unclassified 

Radiological Records facility. Classified documents reviewed included health physics technical 

reports, correspondence, monitoring data, and a limited amount of operational information. 

Two weeks prior to SC&A’s onsite visit, ORAU team members were onsite to conduct site 

expert interviews and records review including classified records. As a result, ORAU has 

recently obtained a large number of documents pertinent to the Y-12 TBDs. In order to prevent 

duplication of effort, Y-12 provided SC&A with correspondence indicating records recently 

provided to NIOSH, and asked that SC&A obtain copies from NIOSH. NIOSH has indicated 

that only documents used in the development of the TBDs had been scanned and placed on the 

O-drive (also referred to as the NIOSH Site Profile Document Database). Most recently 

retrieved data were not available for review by the SC&A team, as they were not yet posted on 

the NIOSH Site Profile Document Database and, in some cases, were “Official Use Only” 

documents. SC&A had limited time to review a portion of these documents during visits to the 

Y-12 Plant. 

In parallel with the classified review, which focused on the validity, completeness, and adequacy 

of data, and the issues brought to the attention of SC&A by site expert interviews, the rest of the 

SC&A team conducted a broader review of the various aspects of the Site Profile, including the 

adequacy and accuracy of film badge data, the analysis of sources of internal dose, etc. This 

review was conducted by SC&A staff who were not privy to classified information and site 

expert interviews, and was not submitted for review by DOE classification personnel. 

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

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

Attachment 6 provides summaries of the interviews conducted by SC&A by teleconference or in 

person in the Oak Ridge, Tennessee, area during the course of this review. The site experts 

included current and former staff from radiation control, operations, environmental monitoring, 

maintenance, and other support organizations. These interviews were conducted during the 

course of the Y-12 site profile review. Each summary is a paraphrase of conversations held with 

a number of site experts, rather than a verbatim transcript. These statements have been grouped 

into categories to provide a linkage with various portions of the Y-12 Site profile. References to 

specific site experts have been omitted for privacy reasons. 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. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

24 of 203 

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 Y-12 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 

substantially impact at least some dose reconstructions. Issues can also be designated as 

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

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

TBDs. In this review, SC&A has identified 10 findings and has not designated any issues as 

observations. 

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 in summary form a list of the issues, 

and to which objective the particular issue applies. 

The TBDs, in many ways, 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. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

25 of 203 

3.0 ASSESSMENT CRITERIA AND METHODS 

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

Y-12 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. 

This review is directed at current TBDs, and does not take into consideration revisions that might 

be forthcoming. For instance, ORAUT-TKBS-0014-5, Revision 1 (Rich and Chew 2005) was 

reviewed; however, based on recent data acquisition efforts by the ORAU team, this TBD is 

undergoing revision by NIOSH. 

 

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 

publicly available relating to the Y-12 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 Y-12. 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. If, for 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

26 of 203 

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

or claimant favorable, this would constitute a technical accuracy issue. 

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 thus 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. In order to 

accomplish this objective, the Y-12 TBDs were compared to the Hanford TBDs, the Savannah 

River Site (SRS) TBDs, and the Iowa Army Ammunition Plant (IAAP) TBD. 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 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

27 of 203 

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%. 

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 of 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 a claimant. 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, 

their supplemental attachments, and TIBs, giving due consideration to the three categories of 

dose reconstructions that the site profile is intended to support. The six Y-12 TBDs provide 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

28 of 203 

well-organized and user-friendly information for the dose reconstructor when adequate data were 

available to do that comprehensively.

ORAUT-TKBS-0014-1, Technical Basis Document for the Y-12 National Security Complex – 

Y-12 Site Profile (Murray 2004a), explains the purpose and the scope of 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. During the course of its review, SC&A was cognizant of the fact 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. As will be discussed later 

in this report, 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. 

ORAUT-TKBS-0014-2, Technical Basis Document for the Y-12 National Security Complex - 

Site Description (Jessen 2004), is an extremely important document, because it provides a 

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

for subsequent Y-12 TBDs. Specifically, this document describes an overview of the processes 

and operations that were conducted at the approximately 531 buildings at the Y-12 site, and the 

associated sources of exposure relevant to dose reconstruction are covered in Attachment B to 

this TBD. 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/sufficient to support dose reconstruction. 

ORAUT-TKBS-0014-3, Technical Basis Document for the Y-12 National Security Complex – 

Occupational Medical Dose, (Murray 2004b), provides a set of procedures for reconstructing the 

radiation exposures of workers from medical radiographic procedures that were required of 

employees at the Y-12 site. SC&A reviewed this section for technical adequacy and consistency 

with other NIOSH procedures, and compared these with the Hanford and SRS site profiles. 

ORAUT-TKBS-0014-4, Technical Basis Document for the Y-12 National Security Complex – 

Occupational Environmental Dose (Ijax and Adler 2004), provides background information and 

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

the facilities at the site who may have been exposed to routine and episodic airborne emissions 

from these facilities. SC&A reviewed this section from the perspective of the source terms and 

the atmospheric transport, deposition, and resuspension models used to derive the external and 

internal doses to these workers. 

ORAUT-TKBS-0014-5, Technical Basis Document for the Y-12 National Security Complex – 

Occupational Internal Dose (Rich and Chew 2005), presents background information and 

guidance to dose reconstructors for deriving occupational internal doses to workers. This section 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

29 of 203 

 

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

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

the recommended assumptions for use in reconstructing internal doses based on whole-body 

counts and bioassay 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. 

ORAUT-TKBS-0014-6, Technical Basis Document for the Y-12 National Security Complex – 

Occupational External Dose (Kerr 2003), 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. 

In accordance with SC&A’s site profile review procedures, SC&A performed an initial review of 

the six TBDs, their supporting documentation, and the six TIBs. SC&A then submitted 

questions to NIOSH with regard to assumptions and methodologies used in the site profile. 

These questions are provided in Attachment 3. A series of conference calls were then conducted 

with NIOSH, ORAU, and the SC&A team to allow NIOSH to provide clarifications and to 

explain the approaches employed in the site profile TBDs. A summary of the conference calls 

with NIOSH, ORAU, Y-12 staff, and SC&A is provided in Attachment 4 and Attachment 5. 

An extensive comparison was done between the methodologies used in the Y-12, Hanford, and 

SRS TBDs to determine occupational medical, environmental, internal, and external doses. This 

comparison focused on the methodologies and assumptions associated with dose reconstruction 

and resultant values used to obtain a POC. A detailed analysis is provided in Attachment 7. 

Information provided in the conference calls with NIOSH was evaluated against the preliminary 

findings to finalize the vertical issues4

 addressed in the audit report. There are three levels of 

review for this report. First, SC&A team members review the report internally. Second, SC&A 

engages an outside consultant who has not participated in the preparation of this document to 

review all aspects of this report. The third level, referred to as the expanded review cycle, will 

consist of a review of this draft by the Advisory Board and NIOSH. The first two levels of 

review have been completed. 

After the Advisory Board and NIOSH have had an opportunity to review this draft, SC&A plans 

to request a meeting with Advisory Board members and NIOSH representatives to discuss the 

report. Following this meeting, we will revise this report and deliver the final version to the 

Advisory Board and to NIOSH. We anticipate that, in accordance with the procedures followed 

during previous site profile reviews, the report will then be published on the NIOSH Web site 

4

 The term “vertical issues” refers to specific issues identified during our review, which were identified as 

requiring more in-depth analysis due to their potential significant impact on dose reconstruction. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

30 of 203 

and discussed at the next Advisory Board meeting. This last step in the review cycle completes 

SC&A’s role in the review process, unless the Advisory Board requests SC&A to participate in 

additional discussions regarding the closeout of issues, or if NIOSH issues revisions to the TBDs 

or additional TIBs, and the Advisory Board requests SC&A to review these documents. 

Finally, it is important to note that SC&A’s 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. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

31 of 203 

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 claims can be processed in a timely manner. With this perspective in mind, 

we identified a number of strengths in the Y-12 site TBDs. These strengths are described in the 

following sections. 

4.1 COMPLETENESS OF DATA 

In an effort to be comprehensive in addressing the range of facilities and processes at the Y-12 

plant, NIOSH effectively compiled facility-specific information on major facilities and 

processes. Descriptions were provided for 17 facilities and 5 time periods, reflecting the 

changing mission over time. A comprehensive and effective overview of key uranium

operations, their location, and the dates of operation is provided in the site profile. SC&A 

considers this to be an important start in providing background information to the dose 

reconstructors. 

In developing the site profile, NIOSH drew upon information contained in 223 reports cited in 

the reference section. These include annual environmental reports, beginning in 1960, that 

present data from offsite releases. Process information was drawn from Oak Ridge health 

studies, the Oak Ridge Dose Reconstruction project, historical overviews, and technical 

documents primarily from public sources. NIOSH/ORAU met with the Y-12 Labor and Trades 

Council on November 9, 2004, to identify worker concerns and discuss the Y-12 TBDs. This 

interaction with workers helps to provide valuable insight into site operations and programs. In 

addition, the issuance of several TIBs and follow-up data captures at the Y-12 facility reflect an 

ongoing effort by NIOSH to continually improve the background information and guidance 

provided to the dose reconstructors. 

4.2 ADEQUACY OF DATA 

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

the Y-12 National Security Complex 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) Historical documentation of the Radiological Control program, including research 

activities, is well documented. 

(3) Evaluations were performed for special projects to identify the radiological hazards 

associated with radionuclides and how they differed from the uranium handled daily. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

32 of 203 

(4) Y-12 implemented environmental monitoring, including stack monitoring, in-perimeter 

monitoring, and offsite monitoring. 

(5) Starting in 1961, dosimeters were coupled with security badging, which helped to ensure 

that all personnel were monitored for external radiation exposure. 

Although Y-12 had significant quantities of personnel monitoring data, as well as field 

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

monitored for external exposure prior to 1961, and support workers were not placed on routine 

bioassay programs, as compared to their operational counterparts. There are also problems with 

the adequacy of data, particularly with regard to neutron dosimetry, early period external 

monitoring, and radiobioassay data. 

4.3 TECHNICAL ACCURACY/CLAIMANT FAVORABILITY 

The Y-12 TBDs exhibited the following strengths in terms of their technical accuracy and 

claimant favorability: 

(1) NIOSH has supplemented the Y-12 site profile documents with a number of 

TIBs that provide further direction to the dose reconstructor. Attachment 2 

lists six technical documents that all have direct bearing on the Y-12 dose 

reconstruction process. SC&A has reviewed these carefully and found each to 

be helpful to the dose reconstructor. These documents are beneficial in 

understanding the application of the six Y-12 TBDs to the dose reconstruction 

process. 

(2) Medical x-ray and photofluorography procedures have been investigated 

thoroughly to determine the radiographic techniques used at the Y-12 Site. 

The Y-12 occupational medical TBD (Murray 2004b) provides tables 

representing four distinct periods of time where doses estimates for individual 

organs are provided for both photofluorographic chest x-rays and 

posterior/anterior chest x-rays. Explicit consideration of photofluorographic 

examinations is especially important because of their potential for relatively 

large exposures, as compared to conventional radiographs recorded on 

photographic film. 

(3) In light of the fact that very little, if any, x-ray output measurements are 

available for the period from 1943 to 1947 at Y-12, ORAUT-OTIB-0006 

(Kathren 2003) has addressed dose potentials and uncertainties from

occupationally related diagnostic x-rays in a generic manner that is also 

applicable, for the most part, to Y-12. Thus, NIOSH has made a credible effort 

in ORAUT-OTIB-0006 to address the technical factors affecting diagnostic 

x-ray dosage, application of technical factors, default values to use during dose 

reconstruction, other factors that can potentially affect patient dose with their 

uncertainties, and an uncertainty analysis that is useful in reconstructing 

diagnostic medical x-ray doses. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

33 of 203 

(4) NIOSH has developed ORAUT-0TIB-0015 (Groer 2004) to introduce the 

Bayesian method for estimating the parameters of a lognormal distribution 

used to impute missing data. The document also presents comparisons of the 

Bayesian approach with estimates obtained using the classical maximum 

likelihood approach. Use of the Bayesian predictive distribution is more 

claimant favorable than the use of a lognormal distribution with fixed pointestimates of the parameters, because the predictive distribution is more widely 

spread than any single lognormal distribution with fixed-point estimates. 

However, we do have some concerns that a simpler and less computationally 

intensive approach could have been developed without sacrificing claimant 

favorability. 

(5) NIOSH has provided a good discussion on solubility of uranium compounds 

and absorption types. NIOSH correctly points out that the absorption type can 

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

(6) NIOSH has provided a good description of in-vitro bioassay methods used for 

uranium at different times at Y-12. NIOSH has developed information on the 

in-vitro bioassay program for the four different eras of operations, which 

facilitates dose reconstructions tailored to the specific isotopes of uranium 

dominant in that era. 

(7) NIOSH has provided a good description of chest-counting methods used for 

uranium and thorium at different times at Y-12. NIOSH points out that 

claimant-favorable assumptions for uranium should be based on conversions of 

93% enrichment for 235U and natural isotopic abundances for 238U. 

(8) NIOSH has put a great deal of effort in developing ORAUT-OTIB-0029 

(Brackett 2005). This TIB allows the dose reconstructor to use all available 

bioassay urine data to derive the intakes based on co-worker data from the Y12 site. However, there are some issues that SC&A feels NIOSH needs to 

address or modify in order for the dose reconstructor to properly derive intakes 

that are more claimant favorable. The methodologies in ORAUT-OTIB-0029 

are only effective if the quality of the input data is found to be adequate. In 

some cases at Y-12, the data quality is sometimes suspect. 

(9) There was a considerable effort to characterize the radiological source terms 

for routine processing of DU, EU, and HEU. The site description outlined 

operations over the period of operations for major Y-12 facilities. Although 

the evaluation of Y-12 operations focused on uranium processing, there were a 

number of operations (e.g. processing of other materials, disassembly of 

weapons) that are not addressed in the site description. 

(10) The external dose assessment methodology considered adjustments for 

changes in dosimetry technology and response functions, and radiation 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

34 of 203 

sources/fields associated with many major workplace locations. Gamma and 

neutron energy spectrums, and corresponding dosimeter responses, are 

explicitly considered in the site profile. 

(11) The documents addressed the external radiation doses to workers in some

detail during the time frame of the 1950s to the 1990s. The external dosimetry 

TBD (Kerr 2003) provides detailed discussions of the historical administrative 

practices, the monitoring techniques and dosimeter exchange frequencies, and 

the dosimeter technology used for beta, photon, and neutron dose 

measurement. 

(12) ORAUT-RPRT-0032 (Kerr 2005a) and ORAUT-RPRT-0033 (Kerr 2005b) 

offer supplemental information that benefits the dose reconstruction process. 

The documents discuss external hazards for particular areas at Y-12. This 

information should be incorporated into the TBD to further support the 

assumptions and methodologies discussed in the TBD. 

4.4 CONSISTENCY AMONG SITE PROFILES 

Although Y-12, Hanford, and SRS all handled uranium, there are differences in the facility 

designs, processes, and radiological practices. In some cases, these differences require sitespecific assumptions in dose determinations. The assumptions developed in the medical 

occupational exposure TBD, as seen in Table A.7-1, Attachment 7, are consistent among the 

sites. In the case of Y-12, Rich and Chew (2005, pg. 6) states: 

...uranium isotopes in various chemical and physical forms have been the 

primary contributors to internal doses to workers...The primary focus, thus, on 

internal dose has been on uranium compounds and alloys over the wide range of 

235U enrichment.

This is also the case with the uranium areas at both Hanford and SRS. All three sites have 

addressed the receipt of recycled uranium and the potential for exposure to impurities in this 

material. The Y-12 TBDs have, however, improved on the discussions found in the Hanford and 

SRS TBDs. In general, the default parameters for input into IREP have remained consistent 

between the TBDs. 

4.5 REGULATORY COMPLIANCE 

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. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

35 of 203 

• 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. 

NIOSH has effectively complied with the hierarchy of data required under 42 CFR Part 82 and 

its implementation guides for monitored workers.
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

36 of 203 

5.0 VERTICAL ISSUES 

SC&A has developed a list of key issues regarding the Y-12 Site Profile. These issues relate to 

each of the five objectives defined in SC&A 2004. Some issues are related to a particular 

objective, while others cover several objectives. Many of the issues raised below are applicable 

to other DOE and Atomic Weapons Employer sites, and should be considered in the preparation 

and revision of other site profiles. 

5.1 ISSUE 1: UNMONITORED SUPPORT SERVICES AND CRAFTS WORKERS 

The TBDs do not address the potential radiation exposure of specific categories of unmonitored 

workers, e.g., support services and crafts workers. TKBS-0014-5 (Internal Dose)(Ijax and Adler 

2005) acknowledges that its Section 5.4 on “unmonitored dose” is “not complete for all 

operations and/or radionuclides.” This TBD speaks to only two instances where “sufficient 

monitoring information might not be available in the worker records:” 1943–1947, where it was 

noted that, “little or no monitoring was performed,” and 1948–1950, where it is noted that data 

for “fluorometric analyses of urine and blood” could not be located. For TKBS-0014-6 (External 

Dose)(Kerr 2003), it is noted that: 

Missed photon dose to Y-12 workers may occur for the following reasons: (1) the 

worker was not monitored before 1961, (2) there is no recorded dose for short 

periods of time after 1961, and (3) the worker’s dose during a monitoring period 

was recorded as zero because the dosimeter response was less than the MDL. 

This TBD goes further and notes that: 

If a worker’s routine duties and work location remained essentially the same 

during the 1950s and early 1960s, it may be feasible to use his recorded annual 

doses in the early 1960s to estimate his missed dose prior to 1961. Methods are 

being investigated at present based on department numbers, job descriptions, and 

work locations that might be used to estimate annual doses for other workers who 

were not monitored for external radiation exposure prior to 1961 and did not 

remain in the same jobs during the 1950s and early 1960s. 

In terms of facility or location information, the TBD notes that: 

Information has not been found that is adequate to describe the potential missing 

photon dose by facility or location within the Y-12 Plant. This is particularly true 

during the early years when the missed photon dose was most significant due to 

the frequent exchange of film dosimeters and the higher MDLs. 

The above limitations would apply to estimations of potential missed neutron dose, as well. All 

of these factors make it unlikely that dose estimations can be made or even bounded because 

monitoring records do not exist for most of these employees, no reliable job or work locations 

are recorded, and they moved frequently between different radiological areas of the plant. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

37 of 203 

Where some limited monitoring or job location records do exist, these records may not be 

reliable, because the monitoring was typically intermittent and not likely representative of the 

range of exposure experienced by workers. Department, building, or work location records are 

not necessarily reliable, because workers indicated that they were listed as being in one 

department or at one location when they actually performed work for other departments or at 

different plant locations. 

5.1.1 Potential Exposure of Y-12 Support Services Workers 

The percentage of unbadged workers ranged from 2%–25% from 1950–1961, respectively, with 

no routine external data from 1948–1950 (no records have been found for the period 1943–

1947). Many of these unmonitored workers were support services personnel, such as janitors, 

steamfitters, maintenance personnel, various skilled crafts, laborers, and guards. Based on 

worker interviews, routine bioassay monitoring among support workers who had access to 

multiple areas of Y-12 radiological operations was not implemented prior to 1994 (when Y-12 

operations were suspended by DOE). Prior to that time, a number of support services and other 

non-production personnel, who may have been monitored via bioassay and in-vivo counting, 

were removed administratively by Y-12 management based on revised action levels and what 

was considered to be “negligible” readings at the time (McLendon 1972). Departments removed 

in 1972 from such routine bioassay included all process maintenance (except Building 9206). 

By the nature of their jobs and work assignments, these workers were routinely exposed to 

highly enriched and depleted uranium, insoluble (“high-fired”) uranium oxides, thorium, 

transuranic contaminants, and various source terms associated with ORNL’s research and isotope 

production operations. Based on worker interviews, janitors and maintenance personnel were 

assigned to clean up HEU solution spills and areas of excess contamination. One janitor 

interviewed noted that he had to wear a respirator just to sweep the steps in Building 9212, which 

were used by production workers on their way to a local break room. Security guards recall 

seeing “green sludge” from piping in some areas of building 9212 during tours. Others indicated 

that they provided janitorial and maintenance support to ORNL operations, including an 

operational area known to contain plutonium in glove boxes. Much of the work was done on offshift periods (after hours) without direct supervision and (according to workers) without always 

following radiological procedures. Work assignments were done on a daily basis and support 

personnel would find themselves in different parts of the plant at any given time, with the 

exception, apparently, of those support workers assigned permanently to certain operational 

areas. For some workers, their work location of record, if it exists, was not necessarily the area 

of the plant where they performed their work. 

While routine exposure records are lacking for many support services workers, maintenance 

personnel permanently assigned to production operations were monitored regularly (starting in 

1954), along with chemical operators and mechanical operators. Their exposure histories are 

instructive as to the potential for intake that existed for even those not directly involved in 

production activities. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

38 of 203 

For example, the Y-12 Quarterly Health Physics Report of March 5, 1963 (Y-12 1963) reported 

that, of the 42 maintenance personnel receiving in-vivo counting, 5 exceeded the Plant Action 

Level (PAL) for uranium or thorium (requiring either removal from the radiological workplace 

or repeating the counting at a later date). The average uptake level determined for maintenance 

workers for that quarter was 57% of the PAL (compared with 80% of the PAL for chemical 

operators and 31% of PAL for mechanical operators). For an earlier quarterly period, the third 

quarter of 1962, 41 maintenance personnel were counted with no one exceeding the PAL’s for 

uranium or thorium, and an average count of 9% of the PAL for uranium and 6% of the PAL for 

thorium. (McLendon 1963).

5.2 ISSUE 2: PRE-1961 EXTERNAL DOSE 

The guiding assumption for assigning pre-1961 external dose for unmonitored workers―that the 

relatively few co-workers that were badged represented the “maximally exposed individuals”―

has not been corroborated by NIOSH. There were several factors associated with the 

Radiological Control (RadCon) program that would influence which workers were badged and 

how the results were recorded. Line supervisors made all badging decisions for groups and 

individuals within groups (Patterson 1957) in an era where production often took precedence 

over safety. Such practices raise doubts over how much weight should be afforded these 

management assignments. Historic discrepancies in recording badge readings also raise 

questions. For example, beginning in mid-1956, all film badges reading “below the minimum 

detectable [were directed to be] recorded as the average of the minimum detectable reading and 

zero, instead of being recorded as the minimum detectable” (West 1956a). In addition, “some 

fraction” of overall Y-12 film badge results were found to be flawed during 1950–1955 due to 

incorrect assignment of badges based on an incorrect determination of whether gamma or beta 

radiation exposure was predominant for a worker. These issues need to be evaluated to verify 

that the “maximally exposed individuals” actually received the monitoring. 

Finally, statistical information on monitored worker populations has been used to reconstruct 

the exposures experienced by unmonitored individuals in a way that is not necessarily claimant 

favorable and does not make provision for considering supplementary information (e.g. from

interviews) in adjusting the statistical parameters used for inferring missing doses. SC&A is 

concerned that there is no adequate explanation or reason to suppose that dose received by an 

individual differs from the monitored population dose by a constant factor. 

5.2.1 General 

The general situation that the TBD presents is that external radiation was not a large issue at the 

Y-12 facility during its years of operations. Internal exposure was the primary concern, even 

after the external monitoring program was established. 

Health physics was preoccupied with [the] internal exposure problem and [was]

not paying too much attention to external monitoring [between] 1956–1960. 

(West 1993) 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

39 of 203 

While the potential for external radiation doses may not have been as large as at some other 

AEC/DOE facilities, the potential for radiation exposure was not negligible at Y-12. For 

example, ORAUT-RPRT-0032 (Kerr 2005a) and ORAUT-RPRT-0033 (Kerr 2005b) give a more 

complete picture of the radiation fields present at Y-12 through the years. ORAUT-RPRT-0032 

(Kerr 2005a, pg. 136) shows a plot of average annual dose verses year from 1952–1978; the 

average was around 170 mrem/year. 

Figure 1: ORAUT-RPRT-0032 (Kerr 2005a) Attachment 1, Page 136 

5.2.2 Tennessee Eastman Corporation

The TBD does not provide a complete historical account for dose reconstruction during the 

period of 1943–1947. This is illustrated by: 

• The number of workers involved is not discussed and is quite large 

• Levels of uranium in air were extraordinarily high at the Calutron operations 

• There is little if any evidence of respiratory protection 

• Individual worker dosimetry was virtually non-existent 

• By the late 1940s, AEC officials and advisors were made aware that uranium workers 

were being excessively exposed to airborne levels of uranium well above the default 

limits meant to protect against radiation set in World War II 

Between 1943 and 1947, the Tennessee Eastman Corporation employed approximately 40,000 

workers, including a large number of women (Herken 2002), to operate the Calutrons and to 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

40 of 203 

recover and process uranium for weapons in Y-12 buildings 9201-1, 9201-2, 9201-3, 9201-4, 

9202, 9203, 9204-1, 9204-2, 9204-3 ,9204-4, 9206, 9207, 9211, and 9212 (Polednak 1982). 

By any scale, the operation there was mammoth. Plans called for installing a 

pair of 500-tank Calutron race tracks end-to-end in twin two-story buildings, 

each measuring four football fields long. The racetracks were on the second 

story; pumps and plumbing for the vacuum system occupied the first floor. 

Logistic and personnel requirements were in proportion. Every pair of vacuum 

tanks required an individual operator seated at a console, continually adjusting 

the current to focus on the beam. An army of technicians was needed to monitor 

the orange uranium-oxide feed material for the beam and later scrape the errant 

green "gunk" -- uranium salts dissolved in carbon tetrachloride -- from the 

insides of each tank. An army of chemists would separate out the silvery white 

powder containing uranium-235 that was left in the receivers following each 

week-long run (Herken 2002).

ORAUT-TKBS-0014-5 (Rich and Chew 2005, pg 23) states: 

From 1943 to 1947 limited monitoring occurred, for which data cannot be 

retrieved. 

West (1993) indicated that there was no external monitoring program for the period of 1943–

1947. NIOSH has not included a methodology for determining either internal or external dose 

for this period of time. Although the Advisory Board voted to recommend a Special Exposure 

Cohort for the TEC years of operation, some methodology for dose reconstruction for non-SEC 

cancers should be included in the TBD. 

5.2.3 Monitoring from 1948–1949 

Weekly film badge monitoring began on a small fraction of the Y-12 population in 1948. Pocket 

ionization chambers (PICs), which are insensitive to beta radiation, under-respond to low energy 

photons, and are not very reliable, were also worn and read daily (West 1993). Monitoring 

during this period of time was focused on Assay Laboratory Radiographic Shop, Spectrographic 

Shop, and “Metal” Machine Shop workers (Kerr 2003, pg. 8). During this 3-year period, the 

percentage of monitored workers increased from about 1% to about 8%. 

The external dosimetry data is not readily available for the period from 1948–1950 for use in 

dose reconstruction. ORAUT-TKBS-0014-6 (Kerr 2003, pg. 8) states the following: 

The external monitoring data for Y-12 workers from 1948 to 1950 are not readily 

available by Social Security Number (SSN) and are not being supplied by Y-12 in 

response to Energy Employees Occupational Illness Compensation Program Act 

(EEOICPA) requests... (Souleyrette 2003) 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

41 of 203 

Given the limited subset of workers monitored and the lack of data available for dose 

reconstruction, the TBD should provide a bounding method for assigning dose for 1948–1949 

assuming one can be developed. 

5.2.4 Monitoring from 1950–1961 

A two-element film dosimeter replaced the PIC as the primary dosimeter in 1950. With the 

introduction of a formal external monitoring program, all personnel working with depleted 

uranium, discrete gamma or beta sources, x-rays, or fission products contaminated materials 

were asked to wear film badges (McLendon 1960). 

Administrative practices for recording dose varied through time, adding to inconsistencies in the 

monitoring program. The program from 1952–1955 differed from the program implemented in 

1950–1951. Instead of assigning a zero to the weekly film badge results less than zero, the 

minimum detectable dose was assigned when a badge result was available. West (1993) 

describes a second administrative practice complicating the interpretation of film badge results 

for the 1952–1955 time frame: 

During this period another practice was followed that further increased the 

difficulty of estimating dose from these data, in that an administrative decision 

was made as to which kind of radiation a person was liable to be exposed and to 

[whether] his/her dose was interpreted as beta or gamma depending on this 

administrative decision. The determined dose or the <MDL assigned dose was 

attributed to this type of exposure. There also seemed to be some change in 

thinking on this matter with years in that in 1952 essentially all cases spot 

checked had all gamma’s and no gamma’s or betas. In 1953, 4 of the 14 cases 

spot checked showed only beta dose. In 1954 and 1955, 13 of the 14 doses spot 

checked showed only beta doses (West 1993). 

Further inconsistencies in recorded data were introduced when West (1956) implemented the 

following policy:

Effective the 22nd week [of] 1956 (May 22 to June 2), all film badges which show 

a reading below the minimum detectable will be recorded as the average of the 

minimum detectable reading and zero, instead of being recorded as the minimum 

detectable. It is felt that this procedure will result in more accurate cumulative 

results on the persons who quite frequently have badge results below the 

minimum detectable. 

Clearly there are uncertainties associated with this data that require further investigation on the 

part of NIOSH if these values are to be used in individual dose reconstructions. The 

administrative decision on whether dose was to be assigned as beta or gamma casts doubt on 

whether a given dose can be appropriately assigned to shallow or deep dose. Review of 

dosimetry processing logbooks may be necessary to determine the usability of this data.
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

42 of 203 

5.2.5 Missed Photon Dose 

Missed photon dose may occur when (1) the worker was not monitored, (2) there is no recorded 

dose, or (3) the workers dose during the monitoring period was recorded as zero when the 

dosimeter response was less than the MDL. The potential for missed dose at Y-12 is greatest 

among workers not monitored prior to 1961 due to the select monitoring criteria. ORAUTTKBS-0014-6 (Kerr 2003, pg. 30) presents methods that may be used to establish missed dose 

prior to 1961. This methodology is further described in ORAUT-PROC-0042 (Kerr and Smith 

2004), which is used by NIOSH to account for incomplete personal monitoring data prior to 

1961. The methodology is based on an extrapolation of the population dose distribution for 

workers monitored from January 1961 to December 1965. A scaling factor is derived for each 

worker having monitoring data in this period. Problems could arise in obtaining a claimantfavorable dose for some workers, such as the following: 

• Workers who were unmonitored and terminated prior to 1961 

• Workers who were unmonitored and switched to low-radiation or non-radiation jobs just 

prior to 1961 

• Workers who performed a number of different jobs, prior to and/or after 1961 

• Workers whose job assignments are unclear or undocumented prior to and/or after 1961 

Appling the methods referred to in the TBD (Watson 1994) requires knowledge of and similarity 

of jobs for the two time periods (prior to and post-1961). Consideration also needs to be given to 

the changes in radiological controls and improved techniques over the course of operation. 

ORAUT-PROC-0042 (Kerr and Smith 2004) provides an alternative method for assigning dose 

under the following conditions: 

• Worker does not have monitoring data for five quarters between January 1961 and 

December 1965 

• Worker was only employed prior to 1961 

• Worker was never monitored for external radiation 

• Worker has evidence of external exposure in their records 

The procedure directs the dose reconstructor to use a scaling factor of 1 and the “population dose 

distributions for monitored workers” (Kerr and Smith 2004). 

The Y-12 policy was to monitor workers expected to receive doses greater than 10% of the 

limits set forth in the radiation protection guides (RPGs) (30 mrem/week photon and 60 

mrem/week beta) (Kerr 2005a). As a result of this policy, the TBD concludes the following 

(Kerr 2003, pg. 8): 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

43 of 203 

Records of radiation dose to individual workers from personnel dosimeters worn 

by workers and co-workers are available for the employees with the highest 

potential for external radiation exposure from 1950 to 1961 and for all workers 

from 1961 to the present. 

During the period of 1950–1961, only certain individuals were included in the film badge or 

external monitoring program. The percentage of workers monitored by film dosimeters 

increased from about 6% to less than 25% between 1950 and 1960 (see Figure 6.1-1 of ORAUTTKBS-0014-6)(Kerr 2003). West (1993) analyzed the number of persons monitored from

computerized and hard copy records for 1950 to 1960. The number of personnel monitored 

ranged from 148 in 1950 to 1,295 in 1960. Starting on June 30, 1961, “nearly all personnel were 

monitored” (Kerr 2003, pg. 8), and the number of personnel monitored jumped four fold (Kerr 

2005a). Information on the change in population dose during this switch, especially by 

department, was not provided. 

Line supervisors made all badging decisions for groups and individuals within groups (Patterson 

1957) in an era where production often took precedence over safety. This practice raises doubts 

over how much weight should be afforded these management assignments. Line supervisors 

were responsible for administering the safety rules and regulations, including (1) informing all 

concerned employees, within the limits of security, of potential radiation hazards and the 

necessary safeguards established to guard against them; and (2) arranging for participation in the 

established personnel monitoring program (Patterson et al. 1957). 

Assignment to Film Program -- Supervision, with the assistance of the Health 

Physics Department, decides which groups and which persons within a group 

shall be assigned to the routine film monitoring program, and keeps current the 

list of persons assigned. 

The supervision makes a request, either oral or written, for the Health Physics 

Department to add persons to or remove persons from the film badge program. 

The Health Physics Department sends the request to the monitoring laboratory 

giving all the necessary data on a "Film Badge Assignment Request" form. 

With monitoring requirements being determined by line supervision, there is no guarantee that 

the monitoring recommendations were consistently implemented. This fact, coupled with the 

lack of concern over external monitoring and the attitude of line organizations towards safety, 

casts doubt on the validity of the TBDs assumption that the highest exposed workers were 

monitored. 

A 1988 appraisal conducted by the Oak Ridge Operations Office (DOE 1988) stated the 

following: 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

44 of 203 

ES&H personnel expressed concern about enforcing compliance issues. Some 

line organizations are either ignoring or are nonresponsive to ES&H 

audits/recommendations. 

This appraisal is indicative of the commitment of line management to the safety program. 

Consistency in the external monitoring program would have occurred when Health Physics took 

responsibility for the issuance of dosimeters. There is no indication in the TBD or supporting 

documents reviewed as to when this occurred. 

In summary, the dose reconstruction methodology utilized for pre-1961 exposures should 

consider the following: 

• The methods used to validate that the highest exposed workers in the work force were 

badged. 

• Consistency in the decisions made by line management to add personnel to the external 

monitoring program, including support personnel. 

• The balance of production verses safety in the workplace. 

• Selective bias due to administrative criteria that assigned badges and type of exposures, a 

priori, is a confounding factor. 

• The methods outlined in the TBD are not claimant favorable if the pre-1961 badging was 

effectively cohort badging of various groups. If cohort badging was widely used at that 

time, the use of the dose distribution for this group would not necessarily be a claimantfavorable surrogate for other unmonitored workers at that time. 

• No documentation has been found describing the criteria used to identify those workers 

who were monitored, such as by random monitoring or temporary 100% monitoring. 

• Maintenance and janitorial employees, and some inspectors and material inventory 

handlers, were not monitored routinely, yet had full access to Y-12 radiological control 

areas. 

• Rovers, whose work included cleaning contaminated areas, fixing contaminated process 

equipment, and moving material, may not have been in a department that was considered 

for monitoring. 

• If records do exist that demonstrate that the individuals that were monitored were selected 

because they had the potential to experience the highest exposures, this information needs 

to be presented in the TBD. 

Each of these issues should be addressed prior to assuming that the worker population that was 

badged reasonably and reliably bounds the exposures experienced by non-monitored workers 

prior to 1961. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

45 of 203 

5.2.6 Statistical Application for the Assignment of Missed Dose 

The purpose of ORAUT-OTIB-0013 (Frome and Groer 2004) is to apply the Bayesian method to 

estimate missing measurements in time periods prior to 1961 for each individual who was 

monitored in at least five quarters from 1961 to 1965. As stated in the TIB, “the method is based 

on the assumption that the individual’s potential for exposure during the 1950s is similar to that 

from 1961 to 1965, and that the individual’s doses differ from the population dose by a constant 

factor.” This factor is the individual scaling factor N described above. This factor is applied in 

each time period to adjust the population dose found in that period either up or down, depending 

on the average relationship between an individual’s doses and the population doses in the time

periods after 1960 when the individual was monitored. Individual doses in each period with 

missing data are simulated in Crystal Ball (an Excel add-in product) using the appropriate 

population lognormal distribution for that period, after the individual scaling factor is applied. 

The distribution is truncated to remain between the 1st and 99th percentile of this lognormal 

distribution. 

The purpose of ORAUT-0TIB-0015 (Groer 2004) is to introduce the Bayesian method for 

estimating the parameters of a lognormal distribution used to impute missing data. The 

document also presents comparisons of the Bayesian approach, with estimates obtained using the 

classical maximum likelihood approach. 

The lognormal model used in ORAUT-OTIB-0015 (Groer 2004) contains four types of 

parameters. The first two are the common shape and scale parameters for the lognormal 

distribution, F2

 and log($), which are constant for all times and all employees. A third type of 

parameter (:t) provides an additional set of scale factors, which are used to model the variation 

in the population for different time periods t. These terms reflect the requirement to account for 

time trends in the pattern of exposure. A fourth type of parameter is used to estimate an 

additional set of scale factors (Nk) for each individual, k, who was monitored in some but not all 

time periods. The variation of individual exposures from general population exposures is 

accounted for using this parameter. The scale factors for individual exposures do not depend on 

time. 

Use of the Bayesian approach permits derivation of an analytic expression for the joint 

distribution for all parameters, given the observed data. This joint distribution reflects the 

uncertainty in the parameter estimates. Use of the Bayesian approach also permits calculation of 

the predictive distribution for a future observation obtained from the distribution estimated in 

each time period for any individual. The predictive distribution is obtained by averaging over all 

possible parameter values, using the joint distribution of the parameters as weights. 

Use of the Bayesian predictive distribution is more claimant favorable than the use of a 

lognormal distribution with fixed point-estimates of the parameters, because the predictive 

distribution is more widely spread than any single lognormal distribution with fixed parameter 

estimates. However, there is a fundamental assumption that the use of observations obtained 

from the time periods when an individual is monitored may be used as a “surrogate” for the 

missing observations in other time periods. This is particularly a concern because in most cases, 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

46 of 203 

the missing observations occur in the earlier time periods when monitoring was less complete, 

using less sophisticated equipment, and operations were conducted with fewer safety controls. 

While this approach may be unbiased based on the available monitoring data, it does not appear 

to be claimant favorable. 

First, the statistical measures shown in Table 1 of OTIB-0013 (Frome and Groer 2004) (μ and σ) 

are treated as true measures of the characteristics of the underlying lognormal distribution, rather 

than estimates. It is instructive to plot these data by quarter. This is shown in Figure 2 below. 

Statistical Measures by Quarter

0

1

2

3

4

5

6

0 10 20 30 40 50 60 70 80

Quarter

Value

m

s

s/m

Figure 2: Statistical Measures by Quarter from Table 1 of ORAUT-OTIB-0013 

(m = μ, s = σ) 

For both μ and σ, we have two distinct periods. The first runs from quarter 1 to quarter 37 

(which is the third quarter of 1956). Over this period, there is a smooth and continuous trend in 

both quantities. This is because the estimates come from a regression analysis on 147 workers 

with dose measurements over the period. The second period occurs subsequently. In this case, 

results are estimated from observations on a quarter-by-quarter basis, and the data exhibit 

considerable scatter. This scatter could be due to two effects; these are actual variations in 

workplace conditions from quarter to quarter and the limited size of the database. The second of 

these should be less important, as most relevant workers should be included, so we are not in the 

situation of extrapolating from a small sample to a much larger population. We, therefore, take 

the view that this variation is due to changes in workplace conditions. At a population level, a 

similar degree of variability should exist before the third quarter of 1956. Therefore, the values 

of μ and σ used for quarters before the third quarter of 1956 are estimates and not the true values. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

47 of 203 

This means that there is an additional degree of uncertainty that is not taken into account in the 

method used. This additional uncertainty should increase the effective σ of the distribution (by 

integrating over the potential distributions in each quarter). However, as can be seen from

examining the values used for σ, these are constant for all quarters before the third quarter of 

1956 and are similar to the typical value after that time. 

Section 3.0 of ORAUT-OTIB-0013 (Frome and Groer 2004) made the following assumption, 

which we believe is flawed: 

The method is based on the assumption that the individual’s potential for 

exposure during the 1950s is similar to that from 1961 to 1965, and that the 

individual’s doses differ from the population dose by a constant factor. 

Our concern is that there is no reason to assume that the dose received by an individual differs 

from the population dose by a constant factor. Over the period of interest, the size of the 

population changed markedly. Figure 3, recreated from ORAUT-OTIB-0013 (Frome and Groer 

2004), which is taken from Watkins et al. (2004), illustrates this point.

Figure 3: Annual Film Badge Monitoring of Y-12 Workers for Gamma Exposure 

(Watkins 2004) 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

48 of 203 

Note the trends both in total numbers of workers and in those monitored through to 1960. Note 

also the jump to complete monitoring of the workforce after 1961. A worker performing a 

specific job with his own particular trend in dose is being compared with a population, which 

exhibits an overall trend in dose, because various individual jobs within that population exhibit 

trends in dose and because the mix of jobs being undertaken will change. In particular, the 

change in monitoring philosophy in 1960 will have brought a large number of jobs into 

monitoring that were not previously represented in the population. In this context, it is of interest 

to note that the first quarter of 1961 is quarter 55 in Figure 2 above. Careful examination of that 

figure suggests an alternative interpretation; with μ having a typical value of around 4.0 between 

quarter 38 and quarter 55, and then jumping down to a lower value of around 3.0 from quarter 

56. From these considerations, it is evident that the analysis makes two jumps of logic: 

• It assumes that the difference between an individual’s dose and the population dose from

1961 to 1965 can be used to infer the difference between that individual’s dose and the 

population dose in the earlier period, even though there was a complete change in 

monitoring philosophy and consequently in the size and variety of work experience of the 

monitored population; 

• It assumes that the time trend in individual exposure mirrors the typical time trend in 

overall population exposure. 

The fallacy of the latter can be readily appreciated by considering an individual who worked 

unmonitored in a relatively high-exposure task in the late 1950s and was then moved to a lowexposure task in 1960. In 1961, he would have been monitored because of the change in 

philosophy. In this case, his post-1961 exposure might be lower than the population average, but 

his pre-1961 exposure might be higher than the average. 

Our review of this protocol reveals that NIOSH attempted to develop a statistically sophisticated 

protocol for filling in missed dose during the early years that, in theory, is statistically valid, but 

in practice, is not necessarily claimant favorable. In addition, when implementing this procedure 

as part of our audit of individual cases, we found the procedure difficult to implement, time 

consuming, and, in the end, not substantially more robust than much simpler, readily 

understandable, and more claimant-favorable approaches to filling in missing dosimetry. 

5.3 ISSUE 3: RADIOBIOASSAY AND INTERNAL DOSE CALCULATIONS 

Potential missed dose associated with inadequate bioassay techniques and sampling delays 

should be further investigated to determine the impact on internal dose calculations. The 

detection limits for uranium prior to 1989 are only established on a provisional basis. There are 

uncertainties regarding the determination of lung counting detection limits for 235U and 238

U lung 

counts that can lead to missed dose. The parameters associated with the intake model used to 

calculate missed internal dose may not reflect actual exposure conditions. These issues must 

also be considered in the development of co-worker data. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

49 of 203 

Several technical accuracy and data completeness issues concerning assignment of internal 

exposures were identified. ORAUT-TKBS-0014-5 (Rich and Chew 2005) is incomplete with 

respect to exposures to radionuclides other than uranium. The site profile does not provide 

information relevant to the calculation of uncertainties in bioassay techniques that were

implemented at Y-12. 

The range of solubilities at Y-12 has not adequately considered the wide range of materials to 

which an individual worker could have been exposed. There is an absence of consideration of 

low-level, chronic exposures by a majority of the population, which is clearly described in Y-12 

historical monitoring documents. Co-worker dose assignments do not consider the shortcomings 

of the uranium urinalysis database, variability in material type, bioassay technique uncertainties, 

and sampling methodology. 

5.3.1 General Information 

The Y-12 internal dosimetry TBD and ORAUT-OTIB-0029 (Brackett 2005) were evaluated to 

determine if adequate guidance was provided to perform a dose reconstruction of internal dose. 

The TBD was assessed with the following criteria: 

(1) Does the TBD provide adequate background information?

(2) Does the TBD provide technically sound assumptions for the calculation of internal 

doses?

(3) Does the TBD provide guidance for addressing missed internal dose for monitored and 

unmonitored workers?

SC&A determined that the internal dosimetry TBD has some deficiencies with regard to 

uncertainty and missed dose (see Issue 3), and does not provide the detailed background 

information and guidance on exposure to radionuclides other than traditional mixtures of 

uranium (see Issue 4). Also, no urinalysis data is available prior to 1948, and data cannot be 

found for 1948–1950. Co-worker data is to be applied to the 1948–1950 time period; however, 

there is no reference to ORAUT-OTIB-0029 (Brackett 2005) that provides information on the 

assignment of co-worker dose. The guidance is incomplete, often confusing, and in some cases, 

appears to contradict guidance provided in other NIOSH documents. 

The Y-12 internal dosimetry TBD (Rich and Chew 2005) is presented in two general sections; 

the main body of the report and Attachment 5A. The main body of the report contains a detailed 

description of the following: 

• The in-vitro minimum detectable activities (MDAs), counting methods, reporting 

protocols, and interferences 

• The in-vivo MDAs, counting methods, and reporting protocols 

• Limited information on the assignment of dose to unmonitored workers 

• Limited discussion of radiobioassay techniques for radionuclides other than uranium
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

50 of 203 

• Discussion of Y-12 uranium solubilities 

There is an extensive discussion on uranium solubility at Y-12. However, there is limited 

direction to the dose reconstructor on the process and assumptions that should be used to 

calculate internal dose. NIOSH has indicated in the TBD that the dose reconstruction

methodology for unmonitored workers for some operations and radionuclides has not been 

completed. Attachment 5A of the TBD contains conventions used in Y-12 internal dosimetry 

reports, general uranium solubility assumptions to be used, in-vivo detection limits, and in-vitro 

detection limits. The attachment is not all inclusive of the directions provided for dose 

reconstruction in the main text. 

The TBD is supplemented by ORAUT-OTIB-0029 (Brackett 2005) and presumably ORAUTOTIB-0002, Technical Information Bulletin – Maximum Internal Dose Estimates for Certain 

DOE Complex Claims (Rollins 2004). ORAUT-OTIB-0029 (Brackett 2005, pg. 3) states the 

following: 

There are instances of energy employees who, for a variety of reasons were not 

monitored for internal exposure during the course of their employment at a U.S. 

Department of Energy (DOE) facility, or whose records of such monitoring are 

incomplete or unavailable. In such cases, data from coworkers may be used to 

approximate an individual’s possible exposure. 

ORAUT-OTIB-0002 (Rollins 2004, pg. 3) indicates that this TIB is to be used under the 

following conditions: 

The purpose of this Technical Information Bulletin (TIB) is to provide a method to 

facilitate timely processing of claims under the Energy Employee Occupational 

Illness Compensation Program Act (EEOICPA 2000), which involve cancer to an 

organ with little or no reported internal dose from internally deposited 

radionuclides that might be associated with work at DOE complex sites. 

There is no reference to these TIBs, which apply to the Y-12 National Security Complex, in the 

internal dosimetry TBD. With such limited guidance provided in the TBD, it is difficult to 

access NIOSH/ORAU methodologies applied to internal dose reconstruction. 

5.3.2 Particle Size 

The TBD has assumed a 5 μm AMAD as the default particle size for Y-12 dose reconstructions. 

Particle sizes differ by chemical form, process temperature, and facility. ORAUT- TKBS0014-5 (Rich and Chew 2005, pg. 10) states the following: 

For different times and different processes, reported particle sizes ranged from 

less than 1 to over 10 μm (physical). Steckel and West (1966) reported a positive 

correlation between uranium oxide particle size and process temperature. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

51 of 203 

Although Y-12 originally assumed a default particle size of 8 μm, the default particle size was 

later changed to 1 μm (Ashley et al. 1992, Snapp 1995). There is no differentiation provided for 

particle size by operation. The TBD has justified the use of a default particle size of 5 μm 

AMAD based on Y-12’s application of the latest ICRP guidance for the lung model, methods 

and models, and weighting factors (Rich and Chew 2005, pg. 10). Furthermore, the TBD states 

the following (Rich and Chew 2005): 

In terms of lung deposition and retention, the 8 μm AMAD class Q material is 

closer to 5 μm AMAD Type M material than to 1 μm AMAD class Y material, as 

shown in Figure 5.3. 

The analysis correlating in-vivo analysis with uranium process materials completed by Steckel 

and West (1966) gives an average particle size range of 0.12 μm for Salvage II operations to 

3.5 μm for burned uranium chips. This average particle size was derived from the portion of the 

process sample that was less than 150 μm, which constituted from 0.2% for Reclaimed Sand to 

99.8% for high-fired UO2. 

The default particle size applied in the TBD conflicts with the direction provided in 42 CFR 

Part 82, which specifically states a default 5μm AMAD particle size is only applicable in cases 

were there is no information on particle sizes available. There were several particle size studies 

completed at Y-12, which indicate that a default 5μm AMAD particle size may not be 

appropriate for all processes at Y-12. 

Although SC&A recognizes that the use of 5μm AMAD particle size is claimant favorable when 

particle sizes are smaller than the default and doses are back calculated from urine results, the 

opposite happens when workers are exposed to larger particle sizes. In this case, workers might 

have had equivalent doses to certain organs underestimated. SC&A recommends that the studies 

conducted at Y-12 concerning particle sizes be reviewed, and the probability of exposure to 

particles sizes larger than 5μm be calculated. The most claimant-favorable particle size 

measured at Y-12 should be considered for each particular cancer type. 

5.3.3 Uncertainties and Detection Limits of the Bioassay Techniques 

The uncertainties in the bioassay techniques and detection limits used to quantify internal dose 

are not fully addressed in the TBD and may affect internal dose reconstruction. 

5.3.3.1 Uncertainties in Bioassay Techniques 

For the early times, NIOSH reports that uranium concentrations in urine samples were 

underestimated. After the change in analytical technique in 1952, reported urinary uranium

concentrations increased. ORAUT- TKBS-0014-5 (Rich and Chew 2005, pg. 23) states the 

following: 

Early in the Union Carbide management at Y-12, the fluorometric method was 

reassessed. Extraction methods were used starting in 1949 and 1950 (CCCC 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

52 of 203 

1949, p. 7; UCC 1950, p. 14). By the first half of 1952, a technique involving 

small (0.2 ml) aliquots of raw urine had been put in place (CCCC 1953, p. 28). 

However, after this change in technique was instituted the urinary uranium

concentrations increased greatly as shown in Table 5.7. This raises concern that 

the technique used before May of 1952 could have underestimated the urinary 

concentrations.

The small volumes of raw urine used for fluorimetry (as little as 0.2 ml) may not be 

representative of a good homogeneous sample, which needs to be taken into consideration in the 

analysis. The principle of using multiple samples, which provides some control on 

homogenization, was not instituted until the mid-1950s. Also, the statement above from

ORAUT-TKBS-0014-5 (Rich and Chew 2005, pg. 23) references Table 5-7, “Typical uranium

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

specifically addresses recycled uranium and does not speak to this issue present above. SC&A 

was not able to identify a table in the TBD discussing the increase in urinary uranium

concentration after May 1952. The lack of needed reference causes confusion. 

In relation to uranium analysis by electrodeposition, ORAUT-TKBS-0014-5 (Rich and Chew 

2005, pg. 23) states the following: 

In 1951, methodological problems leading to underestimates of a factor of two 

were noted and corrected. The information available was not sufficient to 

determine if the records themselves were corrected. 

The lack of confidence in bioassay results was also mentioned. For example, for uranium

analysis by electrodeposition and alpha counting (1950–1989), it is reported in ORAUT- TKBS0014-5, (Rich and Chew 2005, pg. 24) that “precision of any one sample was acknowledged to 

be relatively low.” Furthermore, on page 20, the following is mentioned in relation to urine 

excretion: 

The fraction of the daily void volume was estimated on the basis of the time 

between the sample void and the previous void... The use of the rate method to 

estimate daily urinary excretion contributed to the uncertainty associated with 

any given measurement. 

This practice continued until 1989, when routine samples were collected over a 24-hour period 

(Rich and Chew 2005, pg. 20). At a minimum, the volume of urine voided should have been 

recorded and compared with the typical volume excreted in 24 hours. This approach is not ideal, 

as the dilution of urine depends on fluid consumption. A better and extensively used technique 

is to measure the creatinine content of the sample and to use this to estimate the fraction of

24-hour excretion. 

In addition to the lack of confidence in measurement techniques, there were difficulties with 

contamination in the Building 9995 urine laboratory, which invalidate the results of urine 

analysis during July 1954 (West 1954a). There continued to be contamination problems through 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

53 of 203 

at least September of that year (West 1954b). There is no mention of this contamination issue in 

the TBD or its relative impact on the urinalysis data available. 

5.3.3.2 Detection Limits 

The detection limits for uranium before 1989 are only established on a provisional basis. 

ORAUT- TKBS-0014-5 (Rich and Chew 2005, pg. 21) reports the following regarding 

fluorometry from 1945 to 1989: 

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

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

uncertainty associated with detection limit of a single measurement.

Furthermore, ORAUT- TKBS-0014-5 (Rich and Chew 2005, pg. 21) reports in relation to 

enriched uranium in urine by gross alpha counting from 1945 to 1989: 

Each disk was counted twice (on two different proportional counters) for 30 min 

per count. If the two results from a single disk did not agree within tabulated 

limits, a third count was made (UCC 1966, p.6) and two concordant counts were 

used. If the average results of the two disks from the same sample did not agree 

within specified limits, then two more plates were prepared volume permitting 

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

terms of formal detection limits is not straightforward and no detailed analysis of 

the statistics of this process has yet be found... The LD is provisionally assumed as 

46 dpm/d before 1965 and 25 dpm/d after 1965.

Although the interpretation of potentially censored data of this type is difficult in a statistical 

sense, the use of multiple samples and measurements eliminates the errors arising from

contamination problems and equipment malfunctions. 

ORAUT- TKBS-0014-5 (Rich and Chew 2005, pg. 30) reports the following: 

Due to the uncertainty about actual methods for determining the lung counting 

detection limits, for dose reconstruction purposes and based on a 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. 

These limits of detection are the ones reported for 1959. 

The TBD indicates that no description of the conversion count rate to activity to mass has been 

found (Rich and Chew 2005, pg. 30): 

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 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

54 of 203 

should be based on conversions of 93% enrichment for 235U and natural isotopic 

abundances for 238

U. 

SC&A discovered a letter specifically addressing the in vivo counter conversion factor in 1959 

(UCNC 1959). 

Since the resolution on the in vivo counter is not perfect, the empirically determined 

conversion factor for normal uranium differs for ...[other Types]. The reason is that

some of the 90 keV gammas from the daughter products of U-238 are counted when one 

is counting the 186 keV region. Since the ratio of daughter products of U-238 and U-235 

cannot always be assumed to be constant, the factor for normal uranium shown below 

can only be said to apply to uranium whose U-235 and U-238 daughter products exist in 

the same ratio as in the material on which the calibration was made. 

This information should be retrieved and used to evaluate the adequacy of the early in vivo

counting data. In-vivo counting conversion factors are not trivial and can affect the internal dose 

calculated from such results. 

There are also questions concerning detection limits for radionuclides other than uranium. The 

detection limit for plutonium before 1988 “has not yet been identified” (Rich and Chew 2005, 

pg. 22). 

SC&A finds the document incomplete in terms of retrieving information on limits of detection 

and counting efficiencies. There should be some workers capable of giving reliable information 

on the methods that were used in the 1980s or as early as the 1970s. The document is incomplete 

in terms of providing information relevant for the calculation of uncertainties to the bioassay 

techniques that were used at Y-12. NIOSH/ORAU recognizes that elements of the TBD are 

incomplete in relation to in-vitro and in-vivo detection limits: 

LD values for some historical techniques remain to be identified, and will be 

reported in subsequent revisions as available. (Rich and Chew 2005, pp. 43-44.)

Furthermore, NIOSH/ORAU has continued to capture data on in-vivo counting and will integrate 

this information into subsequent revisions of the TBD (see Attachment 4). 

5.3.4 Interferences 

The selected background interferences to uranium excretion contradict values present by 

Eckerman and Kerr (1999) in their uranium exposure study. The ratios of 234U/238U, which is 

used to distinguish natural uranium, are not realistic. NIOSH is not applying the background 

corrections described in Section 5.2.4, Interferences (Rich and Chew 2005, page 27). As a 

result, NIOSH/ORAU indicated this section would be removed in subsequent revisions of the 

TBD (see Attachment 4). 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

55 of 203 

5.3.5 Potential for Ingestion Exposure Pathways 

The ingestion pathway of internal exposure has not been considered in the TBD. Given the lack 

of contamination controls indicated by radiological control program audits, ingestion seems a 

likely route of internal exposure. In a DOE Oak Ridge Operations audit conducted in 1984 

(DOE 1984), DOE alludes to the potential for ingestion uptakes: 

The act of smoking or consuming foodstuffs with contaminated hands represent a 

clear-cut internal exposure pathway to workers at Y-12. This practice cannot in 

the opinion of the appraiser be considered ALARA. 

Y-12 continued to allow eating, drinking, and smoking in radiological areas until 1988–1989, 

and did not practice egress monitoring (DOE 1986; DNFSB 1993). 

The amount of uranium that enters through ingestion is important in terms of bioassay results 

and doses per Bq excreted. For example, if a worker’s exposure had lasted 1 year, the 1-year 

committed equivalent doses to the different organs, per Bq 234U excreted after the weekend at the 

end of the working year, is illustrated in Table 1. 

Table 1: One-Year 234

U Committed Equivalent Doses per Bq 234U Present in Urine 

after Weekend Urine Sample, Collected at the End of 1-Year Exposure 

F M S

Ingestion

f1=0.02 

Ingestion

f1=0.002 

Sv/Bq 

excreted

Sv/Bq 

excreted

Sv/Bq 

excreted

Sv/Bq 

excreted

Sv/Bq 

excreted

Adrenals 7.69E-05 3.32E-05 2.67E-05 7.59E-05 7.59E-05

Bladder Wall 8.21E-05 3.55E-05 2.84E-05 8.09E-05 8.09E-05

Bone Surface 8.57E-03 3.69E-03 2.96E-03 8.45E-03 8.45E-03

Brain 7.69E-05 3.32E-05 2.66E-05 7.59E-05 7.59E-05

Breasts 7.69E-05 3.32E-05 2.67E-05 7.59E-05 7.59E-05

Esophagus 7.69E-05 3.32E-05 2.67E-05 7.59E-05 7.59E-05

St Wall 7.79E-05 3.76E-05 1.40E-04 1.41E-04 7.30E-04

SI Wall 7.92E-05 4.36E-05 3.10E-04 2.36E-04 1.71E-03

ULI Wall 9.09E-05 9.68E-05 1.74E-03 1.05E-03 9.95E-03

LLI Wall 1.18E-04 2.19E-04 5.04E-03 2.91E-03 2.91E-02

Colon 1.02E-04 1.49E-04 3.16E-03 1.85E-03 1.81E-02

Kidneys 7.95E-03 3.49E-03 2.82E-03 7.84E-03 7.84E-03

Liver 3.51E-04 1.49E-04 1.18E-04 3.47E-04 3.47E-04

Muscle 7.69E-05 3.32E-05 2.68E-05 7.59E-05 7.59E-05

Ovaries 7.69E-05 3.32E-05 2.68E-05 7.59E-05 7.70E-05

Pancreas 7.69E-05 3.32E-05 2.67E-05 7.59E-05 7.59E-05

Red Marrow 9.02E-04 3.87E-04 3.11E-04 8.90E-04 8.90E-04

ET Airways 8.13E-05 6.84E-02 3.05E+00 7.59E-05 7.59E-05

Lungs 1.24E-04 1.29E-01 4.27E+00 7.59E-05 7.59E-05

Skin 7.69E-05 3.32E-05 2.66E-05 7.59E-05 7.59E-05

Spleen 7.69E-05 3.32E-05 2.67E-05 7.59E-05 7.59E-05

Testes 7.69E-05 3.32E-05 2.66E-05 7.59E-05 7.59E-05
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

56 of 203 

F M S

Ingestion

f1=0.02 

Ingestion

f1=0.002 

Sv/Bq 

excreted

Sv/Bq 

excreted

Sv/Bq 

excreted

Sv/Bq 

excreted

Sv/Bq 

excreted

Thymus 7.69E-05 3.32E-05 2.67E-05 7.59E-05 7.59E-05

Thyroid 7.69E-05 3.32E-05 2.66E-05 7.59E-05 7.59E-05

Uterus 7.69E-05 3.32E-05 2.66E-05 7.59E-05 7.59E-05

Years of Exposure to 234U: 1 year

Collection of after weekend urine sample: last month, of the 1st year 

Equivalent Doses calculated for the 1st year after the beginning of work

Excretion of 234 U in urine entirely due to inhalation exposure of Type F material (F) 

Excretion of 234 U in urine entirely due to inhalation exposure of Type M material (M) 

Excretion of 234 U in urine entirely due to inhalation exposure of Type S material (S) 

Excretion of 234 U in urine entirely due to ingestion of soluble material (f1=0.02) 

Excretion of 234 U in urine entirely due to ingestion of insoluble material (f1=0.002) 

As can be seen from Table 1, doses coming from the ingestion pathway should not be ignored, 

especially in the case of cancer to the gastrointestinal (GI) tract. Figures 4 and 5 graphically 

represent the information provided in Table 1. 

 

One Year Committed Equivalent Dose

(Sv per B q excreted in after weekend urine samples after o ne year o f wo rk)

1.00E-05

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

Adrenals

Bladder Wall

Bone Surface

Brain

Breasts

Esophagus

St Wall

SI Wall

ULI Wall

LLI Wall

Colon

Kidneys

Liver

Muscle

Ovaries

Pancreas

Red Marrow

ET Airways

Lungs

Skin

SpleenTestes

Thymus

Thyroid

Uterus

Organs

Com

mitted Equivalent Doses 

(Sv/ Bq excreted in urine)

Inhalation Type F Inhalation Type M Inhalation Type S Ingestion f1=0.02 Ingestion f1=0.002

Figure 4: One-Year Committed Dose in Sv per Bq Excreted in After Weekend Urine 

Samples after One Year of Work 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

57 of 203 

One Year Committed Equivalent Doses

0.00E+00

5.00E-03

1.00E-02

1.50E-02

2.00E-02

2.50E-02

3.00E-02

3.50E-02

Adrenals

Bladder Wall

Bone Surface

Brain

Breasts

Esophagus

St Wall

SI Wall

ULI Wall

LLI Wall

Colon

Kidneys

Liver

Muscle

Ovaries

Pancreas

Red Marrow

Skin

Spleen

Testes

Thymus

Thyroid

Uterus

Organs

S

v/B

q excr

eted aft

er w

eeken

d

Type F

Type M

Type S

Ingestion soluble

ingestion insoluble

Figure 5: One-Year Committed Equivalent Doses 

SC&A recommends that ingestion of uranium compounds be included as a part of the internal 

dose for GI tract. 

5.3.6 Co-worker Dose Assignment 

ORAUT-OTIB-0029 (Brackett 2005) is intended to supplement the internal dose TBD (Rich and 

Chew 2005). This TIB is directed to energy employees who, for a variety of reasons, were not 

monitored for internal exposure during the course of their employment at Y-12, or whose records 

of such monitoring are incomplete or unavailable. In such cases, data from co-workers may be 

used to approximate an individual’s possible exposure. 

Several problems exist with the model used to assign doses to individuals based on co-worker 

results. The database of uranium urinalysis records for the Y-12 site for 1950–1988 from Oak 

Ridge Institute for Science and Education (ORISE), Center for Epidemiologic Research (CER) 

was used without questioning the accuracy of these records. The records were used, although 

there were problems with the database, as stated in ORAUT- OTIB-0029 (Brackett 2005, pg. 3): 

The database results are in units of disintegrations per minute (dpm)/day, 

although original urinalysis results were reported in terms of either mass or 

activity concentrations, depending on the measurement method. The assumptions 

used to convert mass results to activity concentrations for inclusion in the 

database are not known, nor are the assumptions used to normalize spot sample 

results to 24 hours.
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

58 of 203 

5.3.6.1 Choice of the 50th Percentile Intake Rates 

Bioassay results were obtained from the ORISE CER Dosimetry Database, which contains 

uranium urinalysis records from the Y-12 site for 1950–1988 (Brackett 2005, pg. 3). The urine 

records from the database were analyzed by month and were fit into a lognormal distribution. 

The 50th and 84th percentiles were calculated for each month. 

The intake rates were calculated using the Integrated Modules for Bioassay Analysis (IMBA) 

Expert Office of Compensation Analysis and Support (OCAS)-Edition computer program.

Intakes were assumed to be chronic exposures via inhalation. The monthly bioassay results were 

used to obtain a series of inhalation intakes. ORAUT-OTIB-0029 (Brackett 2005, pg.6) advises 

the use of the 50th percentile intake rates to calculate doses. 

There is no explanation on the choice of the 50th percentile. SC&A did not have access to the 

database, and thus could not validate the assumption that the urine activity data is appropriately 

fitted by a lognormal distribution. Monthly urinary results from Y-12 workers from 1952–1988 

were characterized by NIOSH as being typical of a lognormal distribution. As expected with 

data that can be modeled by a lognormal distribution, the urine activity results must be positively 

skewed for each month during the 28 years. SC&A does not agree with the choice of the 50th

percentile to characterize the unmonitored worker’s intake from co-worker data as a claimantfavorable approach. A more appropriate claimant-favorable assumption is to use the upper 95th

percentile urine activity to calculate intake rates and doses. 

The last 10 years of the Table A-1 of the TBD, for example, are characterized by 50th percentile 

values below the detection limits for natural uranium (11 dpm/d), showing that the majority of 

the data were below detection limits. The high value of the geometric standard deviation, on the 

other hand, is a hint to the fact that the positive values probably are much higher than the 

detection limits. The intake rates as derived for March 1, 1978 to September 30, 1984 and from 

October 1, 1984 to December 31, 1988 are very small, and lead to equivalent doses to body 

tissues and organs that are also insignificant. This is clearly not a claimant-favorable approach, 

since many of the workers presented high urine activity concentrations as a consequence of high 

intakes of uranium. Since the high results are consistently measured every month (each month 

results are characterized by the positively skewed lognormal distribution), one can conclude that 

contamination was happening regularly at Y-12. Even for Type S uranium, whose excretion rate 

decreases very slowly, the conclusion stays the same. The contribution of 1 month of chronic 

exposure to the activity of a urine sample collected at the end of the following month does not 

justify the results on Table A-1. The uranium activity of Type S compounds in urine samples 

taken 1 month post-exposure are expected to be 5 to 10 times lower than post-weekend samples 

taken during the month of chronic exposure. Thus the existence of a lognormal distribution 

every month, with the parameters described in Table A-1, leads to the conclusion of regular, 

monthly exposures at Y-12. SC&A considers that the claimant-favorable approach to workers 

not monitored is to consider them exposed to the higher level of contamination that was 

characteristic of Y-12. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

59 of 203 

5.3.6.2 Material Types 

Uranium urine results were fit using Type M and Type S material. Exposures to Type F material 

were not considered, although in ORAUT-TKBS-0014-5 (Rich and Chew 2005), it is specifically 

reported that workers were exposed to uranium compounds of different solubility, including 

soluble material. For example, ORAUT- TKBS-0014-5 (Rich and Chew 2005, pg. 7), reports 

the following: 

The uranium compounds with which Y-12 has worked ranges from highly soluble 

to very insoluble. .... Exposures to soluble compounds were monitored from the 

closing days of World War II by clinical tests of renal function and by fluorimetric 

tests for uranium in urine. 

Furthermore, ORAUT- TKBS-0014-5 (Rich and Chew 2005, pg. 31), states the following: 

From 1948 to 1950, fluorimetric analysis of urine and blood were conducted as 

part of general medical surveillance to prevent kidney damage from exposure to 

soluble uranium compounds; these data cannot be found at this time. Coworker 

data will be applied to this time period. 

This paragraph specifically implies that co-worker data should cover the lack of bioassay data for 

soluble uranium compounds for this time frame. ORAUT-TKBS-0014-2 (Jessen 2005, pg. 8) 

also contains information on exposure to soluble compounds of uranium, which is illustrated on 

Table 2.9-1 (pg. 21). 

The interpretation of results from urinary excretion of uranium Type F materials after 48-hours 

absence from work produce intake rates that result in doses that, depending on the time of 

exposure and organ (primarily systemic organs), can be higher than if Type M or S were used. 

The values presented in Table 2 illustrate these findings. These data are calculated for a worker 

exposed for 1 year to chronic, constant daily intakes. Figures 6 and 7 graphically represent the 

Table 2 Committed Equivalent Dose. 

Table 2: One-Year 234

U Committed Equivalent Doses per Bq 234U Present in Urine 

after Weekend Urine Sample, Collected at the End of 1-Year Exposure 

Type F Type M Type S Dose

comparison 

Dose

comparison 

Sv/Bq excreted Sv/Bq excreted Sv/Bq excreted TypeF /TypeM Type F/ Type M 

Adrenals 7.69E-05 3.32E-05 2.67E-05 2.32E+00 2.88E+00

Bladder Wall 8.21E-05 3.55E-05 2.84E-05 2.31E+00 2.89E+00

Bone Surface 8.57E-03 3.69E-03 2.96E-03 2.32E+00 2.90E+00

Brain 7.69E-05 3.32E-05 2.66E-05 2.32E+00 2.89E+00

Breasts 7.69E-05 3.32E-05 2.67E-05 2.32E+00 2.88E+00

Esophagus 7.69E-05 3.32E-05 2.67E-05 2.32E+00 2.88E+00

St Wall 7.79E-05 3.76E-05 1.40E-04 2.07E+00 5.56E-01

SI Wall 7.92E-05 4.36E-05 3.10E-04 1.82E+00 2.55E-01
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

60 of 203 

Type F Type M Type S Dose

comparison 

Dose

comparison 

Sv/Bq excreted Sv/Bq excreted Sv/Bq excreted TypeF /TypeM Type F/ Type M 

ULI Wall 9.09E-05 9.68E-05 1.74E-03 9.39E-01 5.22E-02

LLI Wall 1.18E-04 2.19E-04 5.04E-03 5.39E-01 2.34E-02

Colon 1.02E-04 1.49E-04 3.16E-03 6.85E-01 3.23E-02

Kidneys 7.95E-03 3.49E-03 2.82E-03 2.28E+00 2.82E+00

Liver 3.51E-04 1.49E-04 1.18E-04 2.36E+00 2.97E+00

Muscle 7.69E-05 3.32E-05 2.68E-05 2.32E+00 2.87E+00

Ovaries 7.69E-05 3.32E-05 2.68E-05 2.32E+00 2.87E+00

Pancreas 7.69E-05 3.32E-05 2.67E-05 2.32E+00 2.88E+00

Red Marrow 9.02E-04 3.87E-04 3.11E-04 2.33E+00 2.90E+00

ET Airways 8.13E-05 6.84E-02 3.05E+00 1.19E-03 2.67E-05

Lungs 1.24E-04 1.29E-01 4.27E+00 9.61E-04 2.90E-05

Skin 7.69E-05 3.32E-05 2.66E-05 2.32E+00 2.89E+00

Spleen 7.69E-05 3.32E-05 2.67E-05 2.32E+00 2.88E+00

Testes 7.69E-05 3.32E-05 2.66E-05 2.32E+00 2.89E+00

Thymus 7.69E-05 3.32E-05 2.67E-05 2.32E+00 2.88E+00

Thyroid 7.69E-05 3.32E-05 2.66E-05 2.32E+00 2.89E+00

Uterus 7.69E-05 3.32E-05 2.66E-05 2.32E+00 2.89E+00

Years of Exposure to 234U: 1 year

Collection of after weekend urine sample: last month, of the 1st year 

Equivalent Doses calculated for the 1st year after the beginning of work

One Year Committed Equivalent Dose

1.00E-05

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

Adrenals

Bladder Wall

Bone Surface

Brain

Breasts

Esophagus

St Wall

SI Wall

ULI Wall

LLI Wall

Colon

Kidneys

Liver

Muscle

Ovaries

Pancreas

Red Marrow

ET Airways

Lungs

Skin

SpleenTestes

Thymus

Thyroid

Uterus

Organs

Com

mitted equivalent dose 

(Sv/

Bq excreted in urine)

Inhalation Type F Inhalation Type M Inhalation Type S

Figure 6: One-Year Committed Equivalent Dose for All Organs 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

61 of 203 

One Year Equivalent Doses

0.00E+00

1.00E-03

2.00E-03

3.00E-03

4.00E-03

5.00E-03

6.00E-03

7.00E-03

8.00E-03

9.00E-03

Adrenals

Bladder W

all

B

one

Surface

Brain

Bre

asts

Esophagus

St W

all

SI Wall

ULI W

all

LLI W

all

C

olon

Kidne

ys

Liver

Muscle

Ovaries

P

ancre

as

Red M

arro

w

Skin

Sple

en

T

estes

Thy

mus

Thyroid

Uterus

Organs

E

quivalent D

oses(

S

v)/

B

q excreted

Type F

Type M

Type S

Figure 7: One-Year Equivalent Doses for Non-Respiratory Tract Organs. 

SC&A finds the assumption that doses should be assigned based on exposures to uranium

compounds of solubility Types M and S, without considering Type F compounds, not claimant 

favorable, for many cancer sites. This assumption does not follow instructions given in 42 CFR 

Part 82, which specifically states the following: 

..., if the solubility classification of an inhaled material can not be determined, 

the dose reconstruction would use the classification that results in the largest 

dose to the organ or tissue relevant to the cancer and that is possible given 

existing knowledge of the material and process. 

The ingestion pathway of exposure was not considered in the derivation of co-worker data. The 

amount of uranium that enters through ingestion is very important in terms of interpretation of 

bioassay results and conversion to organ equivalent doses per Bq excreted. The use of bioassay 

results to back-calculate intake and doses will produce higher doses for certain organs, if the 

ingestion pathway of exposure is used, instead of the inhalation pathway. As seen in Figure 8, 

ingestion of insoluble compounds results in doses to the GI tract that are higher than if inhalation 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

62 of 203 

was assumed. SC&A finds that it is not claimant favorable to ignore the ingestion pathway of 

internal contamination. The ingestion of insoluble compounds should be included as a part of 

the internal dose in cases of GI tract cancers. 

One Year Committed Equivalent Doses

0.00E+00

5.00E-03

1.00E-02

1.50E-02

2.00E-02

2.50E-02

3.00E-02

3.50E-02

Adrenals

Bladder Wall

Bone Surface

Brain

Breasts

Esophagus

St Wall

SI Wall

ULI Wall

LLI Wall

Colon

Kidneys

Liver

Muscle

Ovaries

Pancreas

Red Marrow

Skin

Spleen

Testes

Thymus

Thyroid

Uterus

Organs

S

v/B

q excr

eted aft

er w

eeken

d

Type F

Type M

Type S

Ingestion soluble

ingestion insoluble

Figure 8: One-Year Committed Equivalent Doses from Ingestion of Insoluble 

Compounds

5.3.6.3 Post-Weekend Sampling 

ORAUT-OTIB-0029 (Brackett 2005, pg. 4) states: 

Because of the nature of work at Y-12, a chronic exposure pattern best 

approximates the true exposure conditions for most workers with a potential for 

intakes. 

A chronic constant intake was assumed by NIOSH, and the results were interpreted without 

considering when the samples were taken. 

ORAUT-TKBS-0014-5 (Rich and Chew, pg. 20) confirms this practice: 

...the primary urine collection method was a spot sample submitted Monday 

morning before entering the work area. 

The TBD derived intakes by back-calculated doses based on urine results and by fitting bioassay 

results for different periods. However, the TBD did not consider the 48-hour absence from

work. As a result, the intake rates derived in ORAUT-OTIB-0029 (Brackett 2005) are lower 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

63 of 203 

than those calculated with the same data, but considering end-of-weekend excretion samples. 

The doses to be assigned using these derived intake rates are not claimant favorable. 

The interpretation of bioassay results for chronic exposures of workers should take into 

consideration the day of the week on which samples are taken. In this way, the short-term

components associated with lung clearance will be accounted for, since the early clearance 

component(s) of excretion may introduce a significant difference before and after an interruption 

in exposure, e.g., the weekend. For uranium exposures, the 48-hour waiting period prior to 

submitting urine samples is particularly relevant. 

The TBD did not consider the fact that urine samples were collected after a 2-day absence from

work. The model used by NIOSH assumed that the worker was exposed for a full 7 days per 

week, and did not take into consideration the depletion of uranium in the system after the 2-day 

absence from work. This approach resulted in an underestimate of the uranium intake and 

exposures experienced by the workers. In order to quantify the magnitude of the underestimate 

of dose as derived by NIOSH, SC&A performed a series of dose calculations which take into 

consideration the fact that the urine samples were collected on Monday morning after some of 

the uranium was excreted over the weekend. For example, after 1 year of exposure, the amount 

of uranium excreted in urine using the NIOSH/ORAU intake model as compared to the SC&A 

approach is underestimated as follows: 

(1) For Type F compounds, 4.3 times lower than if samples were assumed to be taken after 

the weekend

(2) For Type M compounds, 2.2 times lower than if samples were assumed to be taken after 

the weekend

(3) For Type S compounds, 2 times lower than if samples were assumed to be taken after the 

weekend 

It is not clear how many working days per year were used by NIOSH/ORAU, and it is likely that 

working schedules of the Y-12 personnel were variable by job title and time period. Even if one 

assumes that NIOSH conservatively considered a 6 or 7 day a week exposure, instead of a 5 day 

a week exposure, this small overestimate of intake will not account for the underestimate of 

intake from the post-weekend sampling. For example, even if exposures were 6 days per week, 

the model will still underestimate the annual exposure by the following: 

(1) For Type F compounds, 3 times for workers on a 5 days a week schedule and 3.7 for 

workers on a 6 days per week schedule 

(2) For Type M compounds, 1.6 times for workers on a 5 days a week schedule and 1.9 for 

workers on a 6 days per week schedule 

(3) For Type S compounds, 1.5 times for workers on a 5 days a week schedule and 1.7 for 

workers on a 6 days per week schedule 

SC&A has included Type F compounds in the comparison, because workers at Y-12 were 

exposed to soluble forms of uranium classified as absorption Type F by the ICRP. As previously 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

64 of 203 

mentioned in this report, Y-12 personnel worked with uranium compounds ranging from highly 

soluble to very insoluble. Intake rates should be recalculated using the fact that urine samples 

were collected after a minimum of a 48-hour absence from the work area. This delayed sample 

collection practice should also be considered for monitored personnel. 

5.3.6.4 Particle Sizes for Co-worker Application 

ORAUT-OTIB-0029 (Brackett 2005, pg. 4) summarizes the breathing rate and particle size 

assumptions for co-workers as follows: 

Intakes were assumed to be via inhalation using a default breathing rate of 

1.2m3

/h and a 5-μm activity median aerodynamic diameter (AMAD) particle size 

distribution.

This value was consistent with the default value recommended by the TBD (Rich and Chew 

2005, pg. 10); however it is not consistent with the particle size data collected by Y-12, as 

discussed earlier in the report. 

SC&A recommends that particle size studies conducted at Y-12 be reviewed, and the probability 

of exposure to particles sizes larger than 5μm should be considered for co-worker data. Workers 

exposed to larger particles might have had equivalent doses to certain organs underestimated, 

when assuming a 5μm AMAD particle size. Although the default particle size is consistent with 

the value recommended in the TBD, this value may not be indicative of the particle sizes 

encountered in all operations at Y-12. 

5.4 ISSUE 4: OTHER RADIONUCLIDES 

The TBD is incomplete in its review of the historic dose contribution of radioisotopes other than 

uranium. These include 3

H, 90Sr, 99Tc, 210Po, 228Th, 232Th, 239Pu, 241Pu, 237Np, 233U, and 241Am. 

Some of these radionuclides were associated with research and development (R&D) activities, 

while others were handled in production, either as source material or as a contaminant from

recycled uranium. While some of these operations are associated with ORNL operations at 

Y-12, a number of the Y-12 workers (e.g., maintenance and janitorial personnel) supported those 

operations and were potentially exposed to these sources. NIOSH has recognized this issue, and 

has directed ORAU to begin evaluating the significance and dose contribution of some these 

materials.

ORAUT-TKBS-0014-5 (Rich and Chew 2005, pg. 7) states the following: 

...although the calutrons in bldg 9204 were left operational for R&D in 

improving the production of the calutron concept and separating other element 

isotopes. Eight calutrons were still operating as late as 1997. This program 

produced an inventory of 225 isotopes from nearly every element in the periodic 

chart. The 86-inch Cyclotron started operation in 1950 and continued to operate 

until 1980...Polonium isotopes and alpha airborne activity are the mentioned 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

65 of 203 

internal dose concerns...There were programs in these R&D efforts, which 

involved plutonium and workers could have plutonium bioassay results in their 

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

research activities involving the Calutron, cyclotron (accelerator), fusion work, 

or plutonium (except in the case of recycle uranium (RU) exposure, which is 

addressed in this section), consideration must be given to possible exposure to 

radionuclides other than uranium. 

In ORAUT-TKBS-0014-5 (Rich and Chew 2005, pg. 12), several radionuclides are listed as 

being radionuclides of concern. These radionuclides are 3

H, 90Sr, 99Tc, 210Po, 228Th, 232Th, 239Pu, 

241Pu, 237

Np, 233U, and 241Am. Other radionuclides cited are 210Po and for organizations outside 

Y-12, 60Co and 95Zr/95Nb. Furthermore, site experts indicated in interviews that radionuclides

processed or worked with at the Y-12 plant included 3

H, 232U, 233U, 237Np, 238Pu, 239Pu, 240Pu, 

228Th, 232

Th, and 241Am. 

5.4.1 Recycled Uranium 

Recycle uranium was received at Y-12 starting in the late 1950s. (see Attachment 8). 

Specifications were established for impurities in recycled uranium; however, this was not done 

until March 1997 (Cox et al. 1997). The health physics division analyzed incoming material for 

106Ru, 144

Ce, 95Zr, transuranics, and other radionuclide concentrations to ensure it was within 

specifications. Health Physics tracked the ratio of impurities to uranium. As recycled uranium

passed through chemical processing, impurities and uranium daughters were removed from the 

product, but concentrated in the raffinate. As a result, the West End Treatment Facility is 

considered a transuranics area. Technetium has become a concern more recently (see 

Attachment 6). 

Overall, the Y-12 TBD contains the most detailed and comprehensive treatment regarding 

exposures to recycled uranium of any TBD reviewed by SC&A to date. It also provides a 

reasonable framework for addressing missed doses from exposures to recycled uranium. This 

approach is a marked improvement over previously reviewed site profiles, where recycled 

uranium was an issue. 

Yet, some important gaps remain that NIOSH should address relative to: (1) process knowledge 

to better inform dose reconstruction; and (2) potential external penetrating radiation exposures 

from recycled uranium in the early period of the Y-12 when external radiation monitoring was 

highly limited. 

5.4.1.1 Lack of Adequate Y-12 Process Information for Recycled Uranium 

The TBD provides default assumptions for internal dose reconstruction based on “the maximum 

values of documented RU levels in the Y-12 process streams and thus represent claimantfavorable values, and which can apply on a plant-wide basis” (Table 5.8 Rich and Chew 2005). 

This approach assumes that the maximum claimant-favorable values can be obtained by 

NIOSH/ORAU from (DOE 2000b). The TBD also had to rely on data on trace contaminants 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

66 of 203 

collected from other sites, such as the Feed Materials Production Center at Fernald, Ohio (Rich 

and Chew 2005). Measurements used in the TBD are suspect because as late as 1985: 

...formal specifications on the maximum quantities of transuranics and fission 

product elements do not exist between the Y-12 plant and feed material 

suppliers...Informal specifications in the form of “gentlemen’s agreements’ did 

evolve and have been used over time... A formal, technically sound, understood 

and accepted specification for maximum transuranic and fission product 

contaminants in uranium recycle material has probably never existed either 

within or between sites... (DOE 1985). 

Documentation or the lack thereof regarding the nature of these informal staff-level “gentlemen’s

agreements,” and their changes over more than 30 years should be addressed by NIOSH. For 

instance, it is quite possible that early staff-level agreements between sites and within the Y-12 

complex were significantly less stringent than later “gentlemen’s agreements.” Transcripts from

the Y-12 Preservation Program, as noted above by the DNFSB staff, in which some 240 Y-12 

employees provided information contained in transcribed interviews between 1993 and 1997, 

may be helpful in addressing this gap in knowledge regarding the changes in recycled uranium

contaminant limits. 

Moreover, it appears that sampling for trace contaminants was not directed primarily toward 

worker protection but rather to ensure successful processing. The DOE’s 1985 Task Force states 

the following (DOE 1985): 

Limited samples exist on the transuranic and fission product content of recycle 

material receipts, processing streams, and product streams. As in the case of the 

FMPC [Feed Material Production Center in Fernald, Ohio], Y-12 has not been 

required to maintain accountability data on plutonium, or other transuranic 

elements, or fission products. 

The Task Force also states the following (DOE 1985): 

...samples are taken primarily for uranium accountability, operational safety, and 

operational control. In general, accountability samples for transuranics are not 

taken for transuranic (plutonium and neptunium) or fission product elements 

introduced, processed or removed from the process. 

In summary form, the current Y-12 recycle material sampling policy associated with 

transuranic and fission product elements is as follows: 

• Sample each batch of receipts 

• If batches received are small, samples from batches are composited 

• Sample one out of every ten of the product batches being returned to the SRP 

• Sample process side streams on a limited frequency (usually annually)

Note these are not accountability samples. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

67 of 203 

Given the very limited sampling of “process side streams,” the values of trace contaminants used 

in the TBD may not be representative of the concentrations encountered in different processing 

stages. Better knowledge of the processing of recycled uranium is necessary to adequately 

inform dose reconstruction. There are some processes where contaminants such as transuranics 

were concentrated, particularly high-heat processing and combustion, which should be 

considered in dose reconstruction. 

In 1985, the DOE task force on recycled uranium noted the following: 

Recent (November 1984) sampling of the S-3 Pond sludge found Pu-238, Pu-239, 

Pu-240 and Np-237 in the sludge...Based on limited sampling since 1977, the 

Y-12 staff has noted a buildup of fission products in both the liquid and solid 

waste streams as a result of processing recycle material. (DOE 1985)

As the following table indicates, dose reconstruction associated with processes where byproducts 

were generated as ash at the Paducah Site indicate significant exposures to transuranics. 

Table 3: Estimated Bone Surface Doses from Recycled Uranium to Workers at the 

Paducah Gaseous Diffusion (Committed Dose Equivalent – CDE in a Year) 

(DOE 2000c) 

Average Air Concentrations Maximum Air Concentrations 

48.06 -- 188 rem 599.24 -- 2,238 rem

A preliminary review of the processes involving the recycle of uranium from chemical 

separations plants and other Y-12 streams indicates that there are several stages where trace 

contaminants could be concentrated in circumstances where workers could be exposed to levels 

higher than reported in Y-12’s sampling data used by NIOSH. Subsequent recovery operations 

involve several steps in which trace contaminants could be concentrated, potentially exposing 

workers. Recovery operations consist basically of the following: 

(1) Burning combustibles―a stage where trace contaminants could be concentrated in ash, 

which may result in worker exposures. 

(2) Dissolving or leaching solids. 

(3) Purifying the uranium-bearing solutions by liquid-liquid extraction. The processing of 

waste streams from primary and secondary extraction stages could also concentrate trace 

contaminants. 

(4) The purified uranyl nitrate hexahydrate (UNH) solutions are denitrated to uranium

trioxide (UO3). This process involves a calciner, which denitrates UNH solution to 

uranium trioxide powder. Nitrogen oxide produced in this process is passed though a 

two-stage scrubber and exhausted to the stack. Maintaining adequate containment of 

radionuclides is critical, as was the case at the Portsmouth Gaseous Diffusion Plant, 

which “ was not able to maintain adequate containment of the radioactive materials 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

68 of 203 

during operating periods” (DOE 1985). Also, bag house residues may contain 

concentrated trace contaminants. 

(5) The UO3 is then reduced to uranium dioxide (UO2) in a fluid bed reactor. Off-gas 

filtration may concentrate contaminants. 

(6) The UO2 is transferred to a hydrofluorinator in which hydrogen fluoride gas is introduced 

to yield uranium tetrafluoride, (UF4) powder. 

(7) The UF4 is then sent to furnaces used to sinter uranium oxide to a ceramic grade U3O8. 

(8) Finally the U3O8 is reduced to metal in a “Bomb Reduction” process. The reaction takes 

place at elevated temperature and reduced pressure in a closed vessel. 

Attachment 8 shows the process flow for uranium at the Y-12 Plant. 

Sources of off-site recycled uranium were not limited to DOE reactors and other uranium plants, 

but also included recovered uranium from commercial reactor fuel. Apparently, recovered 

uranium from the Dresden reactor “B Core” fuel was shipped from the Savannah River Site 

(DOE 1997) and processed at Y-12 (Cox 1997). The Dresden 1 reactor was the first commercial 

nuclear power plant that went on line in 1959. Some of the reactor’s spent fuel was composed of 

irradiated thorium and uranium oxide and was estimated to contain 76.5% 235U, 9.7% 233U, and 

140 ppm of 232U (Cox 1997). 

NIOSH should undertake validation of the secondary sources it has used to ensure that it has 

provided adequate bounding estimates for the highest impurity concentrations in recycled 

uranium.

5.7.5.2 External Exposure from Recycled Uranium 

The TBD does not address exposures to photons and neutrons that are associated with recycled 

uranium, particularly prior to 1961―a period when 85% to 90% (depending on the year) of the 

Y-12 workforce was not monitored for external radiations (Watkins 1994). NIOSH should 

address these potentially significant missing doses. 

In 1985 DOE reported the following regarding worker exposures and recycling at Y-12: 

The exposure records of recycle material workers were compared with those 

within the same operating departments. The results of this comparison shows that 

recycle material workers, on the average, had about 2.7–3.0 times the external 

radiation exposure as other workers in the department. (DOE 1985) 

Since a review done by Y-12 management in 1985 clearly shows a significant increase in 

external doses to recycle material workers, it is reasonable to expect that external doses to Y-12 

recycle workers were likely to have been higher prior to 1961, when far less importance was 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

69 of 203 

given to external radiation by Y-12 management. Only after the criticality accident in 1958 at 

Building 9212, which resulted in very high doses and immediate illness to several unmonitored 

workers, did the Union Carbide Management realize the importance of external monitoring for 

all workers (West 1993). 

Radionuclides of concern, which NIOSH should consider for purposes of external dose 

reconstruction include 103

Ru, 106Rh, 125Sb, 95Zr, 95Nb, and 137Cs, and 144Ce, and decay products of 

232U, which is discussed in more detail later in this section. 

5.4.2 Uranium-233 Production 

Between 1953 and 1961, thousands of kilograms of recycled uranium received from chemical 

separation plants was processed at Y-12 (DOE 1985). Recovered uranium from reactor recycle 

handled at Y-12, in some instances, had 233U contents as high as <98% and ~95%. Uranium-232 

contaminants were as high as 218 ppm (Cox 1997). Of particular concern is exposure to 232

U 

contaminants. Uranium-232 is co-produced with 233U by irradiation of thorium, and is 60 

million times more radioactive than 238

U. This is due to high-energy gamma radiation emitted in 

the decay scheme of 232U daughter products (228

Th, 244Ra, and 228Tl). Even though 232U 

concentrations are small, its gamma radiation constitutes a potentially significant external 

hazard. A 2000 DOE contractor made the following assessment: 

A 50 ppm U-232 content equates to approximately 13R/hr at 1 foot and with 

extrapolation, a 5 to 10 ppm content would emit approximately 5R/hr [at 1 foot].

(Frieboth 2000) 

The implications are that the TBD will need to carefully consider the magnitude of external 

exposures that might have been experienced by unmonitored workers who worked in the vicinity 

of 233U production. 

Y-12 was also involved in processing of 233U; however, no guidance is provided on when 

internal dose from 233U is to be assigned in the internal dosimetry TBD. There is no discussion 

of the potential external exposure from 232U daughters. Also not considered was the potential for 

exposure from salvage material. West and Roberts (1962) describe the 233

U processing as 

follows: 

In the early part of 1962, the Y-12 Plant undertook the fabrication of metallic 

U-233. This was accomplished by chemical conversion of a nitrate solution of 

U-233 to metal and metallurgical fabrication to the desired parts. 

West and Roberts (1962) further discuss the health physics concerns associated with this 

material: 

(1) High alpha activity (the specific activity of the material is about 150 times 

that of the highest assay uranium routinely processed), and 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

70 of 203 

(2) Gamma radiation buildup created by the daughter products of the 

contaminant U-232 accompanying the U-233 

The Y-12 Health Physics groups recognized the special precautions necessary for work with this 

material, including use of containment; prohibiting eating, drinking, and smoking in the 

production areas; requiring personal and clothing surveys with portable instruments upon exit; 

and providing supplemental dosimetry for workers directly involved in processing of material. 

As the material was processed, the 232

U daughters were concentrated: 

Activities in this process concentrated the U-232 daughters in the salvage and 

reduced them in the product material by a factor of about two. Under conditions 

of this processing, this effect was of little significance. However, under other 

operating conditions, this concentration in salvage material could require special 

consideration. (West and Roberts 1962)

In these operations, it was reported that 500 grams of 233U received as uranyl nitrate solution 

contained 40 ppm of 232U (West and Roberts 1962). Although there were special precautions 

taken during processing, it is unclear whether the same considerations were given to salvage 

operations. Potential exposures should be evaluated against the personnel monitoring used at the 

time of exposure. For example, if the feed material was type F material as indicated by West and 

Roberts (1962), the 48-hour delay prior to sampling will impact the activity in the urine sampled. 

The TBD should provide additional information to the dose reconstructor relating to operations 

involving 233U, 232U, and daughters, and provide guidance on when to assume an uptake of these 

radionuclides. The TBD should discuss the extent to which recycled uranium with potentially 

hazardous amounts of 232U may have been processed at Y-12. Furthermore, the TBD should 

consider the impact of exposure to 233U, 232U, and daughters during the handling of salvage 

operations and waste management activities. The analysis should consider time since 

purification. 

5.4.3 Thorium Workers 

The internal TBD provides only a limited discussion of potential thorium and thorium daughter 

exposure as a result of production activities (232

Th) or as a contaminant of recycled uranium

(

228Th). Some discussion of bioassay techniques used at Y-12 for the detection of 228Th, 230Th, 

and 232Th is needed, because thorium was an integral part of manufacturing operations at Y-12. 

There were heavy thorium campaigns in the 1960s, as indicated by the following (Oliver 2001): 

The Y-12 NSC began processing thorium in the early 1960s. Thorium metal in 

pellet form was pressed and/or rolled, formed, and machined. Metal scraps and 

chips were salvaged and also pressed into electrodes to be used in the arc-melting 

process. 

The following primary activities, by facility, occurred as part of this process: 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

71 of 203 

• Pellet/scrap preparation, arc melting, crop and trim machining and sawing 

occurred in 9201-5 

• Mold press sintering, ingot forging and annealing after first cold roll, and 

final inspection/assembly were performed in 9204-2 

• Ingots were canned prior to first cold roll in 9201-1 

• Cold and hot rolling were conducted in 9215 

• Cleaning and final plating were conducted in 9206 

• Machining activities were conducted in 9766 

• Development activities were conducted in 9202 

Thorium is highly radiotoxic because of its long biological half-life and high effective energy. 

West (1965) recognized the health physics hazards associated with thorium production at Y-12. 

These chemical and metabolic characteristics cause it to have a long biological 

half life in the lung and bone. These long half lives, coupled with its highly 

effective energy for internal damage because of the large number of alphaemitting daughters, cause it to be considered a highly radiotoxic isotope. 

Coupled with a long biological half-life that mitigates against effective urinalysis detection and 

the lack of a sufficiently energetic gamma component to permit ready in-vivo detection 

(detection via daughters requires comparison of ratios), thorium presented significant control 

challenges at Y-12 (McLendon 1960; West 1962). 

As the material is processed, thorium daughters are concentrated (West 1965). 

In Y-12 experience, concentration of these less noxious daughters associates itself 

with two situations: (1) there is a concentration of beta and alpha-emitting 

isotopes which come after thoron in the decay chain due to the fact that thoron 

escapes from the parent material because it is a noble gas, and (2) radium 

daughters are separated during operations involving high temperatures. 

Radium-228 and 224Ra exposures can be significant hazards, and internal dose must not overlook 

the potential contribution of dose from these daughters. Although Y-12 evaluated airborne 

concentrations and monitored workers with in-vivo counting, it is uncertain how extensive the 

monitoring program was and how frequently hands-on and support personnel were monitored. 

Although some of the work with thorium was in a contained system, some portions of processing 

(e.g., rolling, machining) were conducted in open areas (West 1965). Machinists handling 

thorium did not take any special precautions grinding and shaping thorium metal (see

Attachment 6). 

Moreover, there was a potential for crossover contamination to uranium production areas. This 

was a serious concern, because Y-12 contamination monitoring techniques (i.e., smearing) could 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

72 of 203 

not distinguish between thorium and uranium contamination (West 1962), nor could it be easily 

picked up in urinalysis. In a February 1963 report, such a “crossover” contamination of plant 

areas with thorium was reported in Building 9201-5. Thorium contamination was tracked out of 

the thorium production areas, leading to a “drastic increase in removable surface contamination 

in the arc furnace area which has resulted in spread to adjacent areas” (Sanders 1963). It was 

noted that this was not a breakdown in equipment, but a problem with procedures and operating 

techniques. Compared with a PAL of 50 d/m/100 cm2

, removable contamination in excess of 

25,000 d/m/100 cm2

 was smeared in the operations area, with 3,000–13,000 and 2,000–

8,000 d/m/100 cm2

, respectively, being smeared in adjacent non-thorium plant areas (e.g., 

beryllium loading dock and Building 9204-4). It is not clear how frequent such crossover 

contamination incidents were or how potential worker exposure to thorium contamination was 

assessed, given the dosimetry challenges cited. As a potential source of missed dose at Y-12, the 

Internal Dose TBD needs to address this issue and characterize its significance for the dose 

reconstructor. 

The TBD should provide additional information to the dose reconstructor relating to operations 

involving thorium and its daughters, including consideration of concentrating daughter products 

during processing and waste management. Chemical and metallurgical processes can displace 

the equilibrium existing in the original source material. Further guidance on when to assign 

thorium uptakes and what default assumptions should be used during various phases of thorium

processing should be provided. 

5.4.4 Technetium and Tritium Exposure 

Although the presence of tritium is acknowledged in the ORAU-TKBS-0014-5 (Internal 

Dose)(Rich and Chew 2005) little is provided characterizing the relative significance of tritium

exposure at Y-12, which categories of work or workers would likely involve potential exposure, 

and whether workers not bioassayed for tritium would have been likely exposed due to 

proximity to tritium sources. Presumably, tritium could have resulted from weapons package 

returns and dismantlement activities, and therefore, workers involved with those operations or 

workers who may have frequented those operational areas may have received exposure and may 

or may not have been bioassayed. 

Technetium is a high specific-activity contaminant whose source is likely recycle uranium feed 

material. 

In a 2001 Y-12 plant assessment of “atypical radionuclides,” the following statement is made: 

A total of 40 samples were taken and analyzed for tritium. The analysis indicated 

that the activity levels were below the laboratory MDAs, i.e., below the 

established values of Appendix D of 10 CFR 835 and DOE Order 5400.5. 

(Y-12 2001) 

A similar conclusion was reached for technetium. However it is not clear, given the passage of 

time and where samples may have been taken, whether this recent survey sufficiently 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

73 of 203 

characterizes historic source terms for these radionuclides and can be used as a basis for 

characterizing past worker exposures to both technetium and tritium. From this standpoint, 

SC&A questions the assertion in ORAU-TKBS-0014-2 (Jessen 2005) that contemporary 

surveys, such as this one, following extensive decontamination activities at the plant over time

and the passage of time itself can be seen as confirming “that the historical specification ensured 

that uranium was the dominating radiological hazard, even against current regulatory 

requirements.” The TBDs need to address the significance of such exposure to determine the 

reliability of what monitoring took place and what the potential for missed dose may have been. 

5.4.5 Neptunium 

ORAUT-TKBS-0014-2 (Jessen 2005) notes that the “primary sources of transuranic materials 

(TRU) [238-240Pu and 237

Np] were from the SRS and the Idaho Chemical Processing Plant (ICPP). 

However, the SC&A review indicates that Y-12 fabricated 237Np alloy rods in specific 

campaigns in the 1960s for which radiation measurements of the rod containers were found to be 

about 250 mR/hour gamma 2 inches from the container (about 15 times that from uranium metal) 

(Francke et al. 1966). Y-12 also apparently considered and may have actually processed 

neptunium in the mid-1990s in Building 9203 (Thomas 1995). Given that the relative 

radiological hazard of 237

Np was found to be about 250 times that of Y-12’s uranium production 

material (Francke et al. 1966), there is a need for the TBDs to better characterize which workers 

were involved in these specific processes and how they were monitored. 

5.4.6 Plutonium and Other Radionuclides 

The TBD addresses plutonium as a radionuclide of concern in the ORNL Y-12 facilities and as a 

contaminant of recycled uranium. Plutonium was not limited to recycled uranium and research 

activities. In fact, there were processes involving plutonium that plant managers denied for 

many years. The TBD assumes Y-12 employees did not support the ORNL operations involving 

radionuclides other than uranium. This was not the case for support organizations, such as 

security, janitorial services, maintenance, etc. (see Attachment 6). 

Plutonium contamination was also a concern regarding salvage beryllium from Rocky Flats 

returns (West 1961). It was necessary for Y-12 to establish an acceptance level for plutonium

contamination of 3.7 x 10-2

 grams Pu/gram beryllium.

The following statement was made regarding processing transuranics at the Y-12 plant: 

Additional, various short-term projects associated with component fabrication 

were performed at Y-12. Some of these involved transuranic material although 

they were not related to the processing of recycled uranium. These operations 

were performed in Buildings 9202, 9205, 9212, and the 9215 M-Wing Machine 

Shop. (Oliver 1997). 

Furthermore, the Health Physics group published uranium and plutonium urinalysis reports by 

building to keep line supervision informed of the various analyses made, including urinalysis and 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

74 of 203 

air sampling data. A review of several of these reports indicated that a limited number of 

personnel from Buildings 9204-3, 9205, and the Special Materials Testing group were monitored 

for plutonium (West 1954c; West 1956b; Wollan 1954). There were also plutonium air sampling 

results from these areas. The inclusion of personnel from these areas in a plutonium bioassay 

program is indicative of work with plutonium in these areas. Plutonium was separated in the 

Calutrons, and was handled at the “plutonium laboratory” and Building 9203 (see Attachment 6). 

Historically, some parts or components were “sputtered” with plutonium. 

The internal dosimetry TBD and dose reconstruction procedures instruct the dose reconstructor 

to assign internal dose from plutonium and other transuranics under the following conditions 

(Rich and Chew 2005): 

(1) When bioassay data is available for the radionuclide, and 

(2) When work involved handling of plutonium such as in Research and 

Development (R&D)

Although information is provided on impurity concentrations in incoming recycled uranium,

there is no direction provided to the dose reconstructor on when internal dose should be assigned 

for these impurity radionuclides. There is no mention of potential missed dose by individuals 

who entered plutonium production and testing areas but were not included in the plutonium

bioassay program (e.g., support services personnel). 

There is no further mention regarding the process and possible exposures in the R&D of the 

Calutrons in Building 9204-3 or exposures in the production of the 225 isotopes produced. 

There is no information on exposures to polonium isotopes and alpha airborne activities. There 

is no description of the work involved and exposures to plutonium, except for the case of 

recycled uranium exposure. There is no further mention or description of possible exposures to 

radionuclides in research activities involving the Calutron, cyclotron (accelerator), and fusion 

work. There is no description of the work involving exposure to 232

Th or 3H, or what is meant by 

“for organizations outside Y-12, 60Co, and 95Zr/95Nb.” 

In summary, SC&A finds that the TBD does not adequately characterize the potential exposure 

to impurity radionuclides in recycled uranium. Consideration was not given to the concentration 

of these impurities and subsequent disposal of the raffinate in the Y-12 waste stream. Potential 

exposure from production operations with 233U, thorium, plutonium, and other radionuclides has 

not been considered. 

5.5 ISSUE 5: NEUTRON DOSIMETRY AND EXPOSURES 

Neutron dosimetry is considerably more complex and difficult to assess than beta/photon 

dosimetry. Principle difficulties in assessing neutron dose relate to the technical capabilities of 

past dosimeters used at Y-12. Initially, Nuclear Track Type A emulsion (NTA) film was 

introduced at Y-12 in 1948 and used through 1989. From 1980–1989, it is assumed that Y-12 

used the same neutron monitoring system as ORNL. This system used a combination of NTA 

film and thermoluminescent neutron dosimeters (TLNDs). It is unclear from the TBD during 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

75 of 203 

what overlapping years both NTA film and TLNDs were used for neutron monitoring at Y-12. 

Since 1989, neutron dose has been measured using an albedo-type dosimeter. 

SC&A has reservations about the assumptions made regarding the ability of NTA film to 

adequately characterize Y-12 worker exposure. Neutron radiation sources are not adequately 

defined in the TBD for all potential neutron exposure conditions at Y-12, including spontaneous 

fission neutrons, moderated (alpha, neutron) sources in solutions/compounds, subcritical and 

critical assemblies, and moderated neutrons from the 86-inch cyclotron. The assumption that 

most neutron exposure at Y-12 was from neutron energies greater than 500 keV, and therefore 

would be detected by the NTA film, is not adequately substantiated. In fact, Wilson et al. (1990) 

state that NTA is a poor detector of film energies between 500 keV and 800 keV. Fix et al. 

(1997) state that the threshold is about 700 keV and may be as high as 1 MeV. In general, the 

NTA film response decreases logarithmically as the energy decreases from 1,000 keV to 500 

keV. Therefore, NTA may seriously underestimate the true neutron dose, and appropriate 

correction factors for this underestimate should be applied. 

5.5.1 Neutron Energy Spectra Characterization 

ORAUT-RPRT-0033 (Kerr 2005b) indicates that occupational exposure to neutrons was largely 

confined to the years 1952–1962 and to certain departments. During this period of time, 143 

individuals have had positive neutron dose at Y-12. Furthermore, the TBD states the following 

(Kerr 2003, pg. 32): 

It was also noted previously that there are a lot of recorded zeros in the neutron 

dose data for Y-12 workers for two reasons: (1) a worker’s NTA film was not 

developed and read, or (2) a worker’s NTA film indicated a neutron dose 

equivalent that was less than the film’s MDL, approximately 50 mrem. 

Prior to 1962, the NTA film had a detection limit of 100 mrem. From 1949–1957, the badge 

exchange rate was biweekly, changing to monthly starting in 1958. By 1961, badges were 

exchanged quarterly (Kerr 2005b). Site experts indicated that when neutron badges were not 

read, the dose was recorded as a zero in the dose of record. The higher detection limit and the 

frequent exchange rates in the early years, and the NTA threshold of around 700–800 MeV, cast 

suspicion on the ability of early neutron dosimetry to effectively measure neutron exposure. 

An adequate characterization of the neutron exposure conditions in light of the limitations of the 

neutron monitoring program should be further evaluated in order to insure the assumptions that 

neutron exposures were confined to particular years, departments, and energies are validated. 

Neutron spectrum measurements are limited to measurements made by PNL of the Calibration 

Laboratory in Building 9983, the Enriched Uranium Storage Area in Building 9212 B-wing, and 

the Nondestructive Assay Laboratory in Building 9720-5. ORAUT-RPRT-0033 (Kerr 2005b, 

pg. 12) further elaborates on the potential areas of exposure to neutrons, including areas in the 

following buildings: 

• Assay Laboratory (9203 Room 8, 9205) 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

76 of 203 

• Criticality Experiments Facility (critical assembly and reactor research)(9213) 

• Electromagnetic Research (9201-2, 9204-3) 

• Health Physics (9983 Calibration Laboratory) 

• Instrument Department (9737) 

• Chemical Operations (9202, 9206, 9212) 

• Cockcroft-Walton linear accelerator

• 5-MeV Van de Graf accelerator. 

NIOSH should also addresses potential exposures from neutrons associated with recycled 

uranium, as indicated in the following statement: 

Neutrons of approximately 2 MeV energy are generated by the interaction of 

alpha particles from uranium with the nuclei of fluorides and other low-Z atoms. 

The magnitude of neutron flux will vary based on the total activity of uranium 

(which is a function of enrichment, U-233 and U-232 content) and the chemical 

compound in question (mixing U and F)... As a result, this concern is primarily 

associated with UF4 and UF6 or with materials containing beryllium, lithium or 

aluminum. (Cox 1993)

In 1997, it was proposed at Y-12 that limits be placed on recycled uranium contaminants, so as 

to reduce the risks of neutron exposures (Cox 1997). NIOSH should determine the extent to 

which the risks of exposure to neutrons, particularly thermal neutrons, from recycled uranium

have occurred, because of the presence of contaminants and inherent process chemicals that may 

have served to enhance neutron flux.

Other areas not considered in the site profile for potential neutron exposure include plutonium

handling areas and other areas where highly enriched uranium may have been stored or 

processed outside the 9212 B-wing. The potential of exposures from lower-energy fission 

neutrons emitted from uranium and transuranic material was also not sufficiently addressed in 

the TBD. This is a potential area of missed or underestimated dose, because exposures from

fission-energy neutrons could result in neutrons below the threshold of NTA film. 

 

The TBD should be augmented to include additional information outlined in ORAUT-RPRT0033 (Kerr 2005b) and site expert input, which provided valuable information of the flow 

process of material from receipt onsite to shipment offsite. Although the TBD has likely 

addressed the highest potential exposure scenarios from neutron, there is further investigation 

required to ascertain whether unmonitored neutron exposure occurred for individuals not directly 

involved in hands-on operations of neutron-producing material. 

5.5.2 NTA Film Response 

The TBD recommends the use of NTA film to assign neutron exposure based on neutron spectral 

studies of limited areas known to have neutron sources. The Hanford and Savannah River Site 

TBDs clearly reject the use of NTA film, based on the neutron energy spectra and the relative 

response of the neutron dosimeter. The focus in these site profiles was to apply a facility-
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

77 of 203 

specific neutron-to-photon ratio to the penetrating dose to obtain a neutron exposure. These 

neutron-to-photon ratios were derived from the same NTA dosimeter information classified as 

inadequate for use in determining neutron dose. The Y-12 TBD abandons this approach, stating 

that the NTA film did not underestimate the neutron exposure, and thus does not need adjustment 

for under-response. 

The recent PNL measurements also indicate that the past NTA film dosimeters 

worked reasonably well in the Y-12 workplace because Ra-Be and Po-Be neutron 

spectra used to calibrate them were reasonably well matched to the workplace 

neutron spectra. These measurements suggest that the NTA film dosimeters 

missed less than 10% of the neutron dose equivalent at the Calibration 

Laboratory of Building 9983 and less than 5% of the neutron dose equivalent at 

the Enriched Uranium Storage Area of Building 9212 and the Nondestructive 

Assay Laboratory of Building 9720-5. (Kerr 2003, pg.27). 

The TBD states that the neutron-to-photon ratios for the Calibration Laboratory, Enriched 

Uranium Storage Area, and the Nondestructive Assay Laboratory are 8:1, 1:1, and 25:1, 

respectively (Kerr 2003). Furthermore, the TBD states that neutron-to-photon ratios can only be 

used for the Enriched Uranium Storage Area in Building 9212. The TBD indicates that the 

neutron-to-photon ratios for the Calibration Laboratory and the Nondestructive Assay 

Laboratory, when applied to the photon dose, would overestimate the neutron exposure. 

The TBD assumes that 90–97% of the neutrons were above a step-function threshold of 500 keV 

(or even above a 700-800 keV threshold), and were detected by the badge. This implies that 90–

97% of the neutron dose is read and recorded in the worker’s dose file. The threshold assumed 

in the TBD conflicts with the accepted threshold of >700 keV used in other TBDs, such as for 

the Hanford site, and may be as high as 1 MeV (Fix 1997). ORAUT-TKBS-0014-6 (Kerr 2003, 

pg. 17) acknowledges that there are limitations in NTA film with respect to low-energy neutrons. 

In general, the response of the NTA film decreases with decreasing neutron 

energy that are greater than a threshold energy estimated to be about 500 keV... 

Results reported at the first AEC Neutron Dosimetry Workshop in 1969 indicated 

that laboratory dose measurements made with NTA film were about one-half to 

one-fourth of those measured with other methods including the TLND. . . The 

response of both dosimeters is highly dependent upon the neutron energy spectra, 

and both dosimeter types require matching the laboratory calibration neutron 

spectra to the workplace neutron spectra for reliable results. [Emphasis added.] 

(Vallario et.al. 1969) 

Given the potential underestimation of neutron dose, an attempt should be made to determine the 

expected neutron energy spectra under actual workplace conditions and in a variety of situations, 

and then apply the appropriate correction factors for each major work area in order to adjust the 

recorded neutron dose to the most likely neutron dose actually received. The amount of neutron 

spectral data is limited to a few areas, which were available for analysis during 1990. This does 

not necessarily represent neutron exposure conditions at earlier time periods, given the change in 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

78 of 203 

material types and quantities and the level of operations prior to the PNL study. Neutron energy 

spectra shift fairly rapidly towards lower energies as surrounding material is added, especially 

hydrogenous material, ultimately affecting the response to the NTA film. 

The TBD states that there are uncertainties in the details of the actual neutron dosimetry program

from 1980–1989 prior to the complete implementation of the TLND. This coincides with higher 

production rates observed in the 1980s. There are implications that Y-12 used the same

dosimeter as ORNL. This neutron dosimeter contained both NTA film and a TLND (Berger and 

Lane 1985). With individuals wearing both types of neutron dosimeters over the same exposure 

period, the NTA film response can be compared to the TLND response, which is capable of 

measuring both slow and fast neutron exposure. If the assumptions made in the TBD that the 

neutron spectra was composed of energies above the NTA threshold, the results from both 

dosimeters would be nearly the same. Further investigation on exactly what the monitoring 

program at Y-12 entailed during these years and subsequent comparison of NTA film and 

TLNDs is a method of comparison that should be investigated to determine the extent of the 

possible under-response in NTA film at Y-12. 

SC&A concludes that NTA film dosimeter data might be deficient, resulting in under-estimation 

of true neutron exposures. Therefore, the use of such data for dose reconstruction may not be 

claimant favorable without validation of NTA results with TLND and field survey instrument 

results. The reference of a 500 keV threshold value for NTA film is not consistent with other 

technical documents, which state NTA film is a poor detector of neutron energies between 500 

and 800 keV (Wilson et al. 1990). Fix et al. (1997) state that the threshold is about 700 keV, and 

may be as high as 1 MeV. The TBD’s reference that more than 95% and 97% of the neutron 

dose is above the NTA film threshold is misleading, since it “suggests” that the film’s 

response/accuracy is 95% to 97%. The response of NTA to neutrons between 500 keV and 

1,000 keV varies by 2 orders of magnitude, as seen in Figure 6.3.2.2-1 (Kerr 2003, pg. 17). The 

implication is that there is a 95–97% response of the NTA film in this energy range, which is not 

the case. 

Further evaluation of the energy spectra is required to validate the assumption that correction 

factors of 1.10 (i.e., 90% of the neutron dose was recorded) for workers in the Calibration 

Laboratory and 1.05 (i.e., 95% of the neutron dose was recorded) for workers in the Enriched 

Uranium Storage Area of Building 9212 and the Nondestructive Analysis Laboratory are 

sufficient to account for under-response in the NTA film. Additional correction factors for areas 

with potential neutron exposure should be developed and applied as appropriate. 

5.6 ISSUE 6: RADIATION GENERATING DEVICES (RGDS) 

The TBD intermittently mentions the use of radiation generating devices at the Y-12 plant; 

however, it does not take into account all radiological hazards associated with the operation of 

these units. Also, the TBD does not discuss how the film badge response to low energy x-rays 

may impact the recorded dose, and what correction factors are necessary to compensate for lowenergy photon exposures from x-ray fields. Although the TBD recognizes that the 86-inch 

cyclotron was used for isotope development, there is no discussion on what the isotopes of 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

79 of 203 

concern for internal dose may be from this operation. Finally, there is no discussion on smaller 

cyclotrons that may have operated for only a short period of time. Although some of these 

devices may have been operated by ORNL personnel at the Y-12 complex, there is still a 

potential for exposures to Y-12 personnel, which should be included in the TBD. 

Radiation generating devices are briefly discussed as a source of radiation exposure in ORAUTTKBS-0014-2 (Jessen 2005, pg. 18). The TBD lists linear electron beams, electron beam 

welders, scanning electron microscopes, x-ray photoelectron spectrometers, secondary ion mass 

spectrometers, enclosed beam diffraction equipment, and x-ray equipment. Energy emitted 

ranged from 15 keV to 9 MeV. ORAUT-TKBS-0014-5 (Rich and Chew 2005, pg. 6) discusses 

the use of a proton-accelerating cyclotron used from 1950–1980 in Building 9201-2. This unit 

produced an inventory of 225 isotopes over the course of operations creating the potential for 

internal exposure to polonium, airborne alpha activity, and other possible radionuclides. 

Although the TBD indicates that the dose reconstructor should consider the potential of intakes 

from radionuclides other than uranium, no additional guidance is provided. ORAUT-RPRT0033 (Kerr 2005b) also provides information on the 86-inch cyclotron, which was a significant 

source of neutron exposure at the Y-12 site. The information from ORAUT-RPRT-0033, which 

provides further information on the role of and radiological hazards associated with the 

cyclotron, has not been incorporated into the TBD and would benefit the dose reconstruction 

process. Furthermore, there are other accelerators that are not mentioned in the TBD that pose 

both internal and external exposure potentials. 

5.6.1 X-ray Machines 

Industrial x-ray units and other potentially high-exposure sources were used at Y-12 for 

nondestructive testing and instrument calibration. Among these units were 10 x-ray machines 

that were in use in this area, ranging in size from 40 kVp, 16 mA to 1 MeV, 10 mA

(Emerson 1951). Struxness (1949) states that a 1 MeV, 3 mA unit was installed in Building 

9212 during January 1949. Center (1950) indicated that x-ray fields were also an issue with the 

high-voltage rectifiers used in many systems. Other devices that could have generated x-ray 

fields included linear electron beams, electron beam welders, scanning electron microscopes, 

x-ray photoelectron spectrometers, and secondary ion mass spectrometers. 

In addition to routine exposures from x-ray units, there were accidental exposures from x-ray 

units, resulting in substantial doses. In 1961, an x-ray overexposure occurred when a worker in 

Department 2259 was exposed to an x-ray beam without realizing the machine was in operation 

at the time. The worker received 4,646 mrem of exposure, 93% of the Radiation Protection 

Guide at the time (McLendon 1962). On April 25, 1965, two employees were inadvertently 

exposed to an x-ray beam while attempting to radiograph a welded part. The machine 

experienced a short circuit, and the unit began operating continuously although the unit appeared 

to be shut off. When a third employee entered the room with an audible pocket radiation 

monitor, the two operators discovered the unit was on. The resulting whole-body doses to the 2 

men were 19 and 11 rem. Portions of the one worker’s hand received an estimated dose of 

3,500 rem (AEC 1965). There are clear situations when dealing with x-ray or other collimated 

units where portions of the body could have experienced significantly higher doses than those 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

80 of 203 

recorded on the film badge. The TBD should explore the potential exposure condition during 

work with these types of units, especially for inadequately monitored or unmonitored workers. 

In addition to the x-ray-producing devices, the site used both portable and stationary gamma and 

photon sources at the Biological Sciences facilities (Dickens 2003): 

• 1,300-Ci and a 65 Ci 137Cs source at Bldg 9207 

• 80-Ci 137Cs source at Bldg 9779-2 

• 3.9-Ci 252Cf (800 mrad/h at 1m, neutron + gamma) at Bldg 9983-17 

A 10-Ci Cobalt source was installed at the Y-12 complex in January 1949 (Struxness 1949). 

The source of radiation exposure from these units would be from the x-ray and gamma photons 

themselves, ranging in energy from 40 kVp to 1.33 MeV. The 2 MeV neutrons from the 

relatively large 252Cf source could also have potential for exposure to workers. The TBD should 

include discussion on the relative response of dosimeters from radiation emissions of this type 

and correction factors for under-response in the dosimetry system. For example, there is no 

indication in the TBD whether film badges were calibrated to match the x-ray energies emitted 

by x-ray equipment. Adjustments for low-energy photon exposure, similar to x-ray exposures at 

Y-12, was a part of the program at Hanford and resulted in an increase in the recorded deep dose 

by applying a fraction of the open window measurement to the deep dose. At the Savannah 

River Site, a special calibration curve was applied to badge results for dosimeters potentially 

exposed to low-energy photons, such as workers in plutonium areas or medical x-ray technicians. 

5.6.2 Accelerators 

Several accelerators were in use at the Y-12 plant. Many of these accelerators were used by 

ORNL at Y-12 for research and development activities and isotope production. Accelerators 

included the 86-inch cyclotron, the 63-inch cyclotron, the 22-inch cyclotron, and the Van de 

Graff accelerator. 

The subject of Y-12 personnel exposure to radiation produced by the 86-inch cyclotron is not 

mentioned in TKBS-0014-6 (Kerr 2003). ORAUT-RPRT-0033 (Kerr 2005b) provides a more 

detailed description of the 86-inch cyclotron as a potential source of neutron exposure, as 

follows: 

Much of the neutron dose to workers at the Y-12 Plant in the 1950s appears to be 

associated with research activities at the 86-Inch Cyclotron. The neutron-togamma dose ratios may have been quite large in the stray neutron fields in 

normally occupied areas near the cyclotron, but they were probably moderated by 

additional exposures to gamma rays from neutron activated materials in the 

cyclotron vault. 

The 86-inch cyclotron was operated from 1950–1980s in Bldg. 9201-2 at Y-12 by ORNL. This 

radiation source created slow and fast neutrons (Center 1950). As of 1950 this unit was 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

81 of 203 

accelerating protons up to 26 MeV at up to 5 mA of current to study (p,n) reactions on copper. A 

neutron flux of 106

 n/cm2/sec was measured inside the pit with 5 ft of concrete wall and ceiling 

shielding around the pit, with two maze openings. Health physics problems were encountered 

from the zinc activation in the copper targets with a 50-microampere beam for a few seconds, 

causing 20 R/hr readings, but with a short half-life. Therefore, the targets were allowed to cool 

and were handled with lead gloves, with plans to install remote handlers (CCCC 1951). The 

isotope production program caused the most exposure. Additional shielding was added around 

door/maze and refrigerant pipe penetrations (Struxness 1953). Each worker in the vicinity of this 

unit was assigned a film badge, two pocket ionization chambers sensitive to gamma and two 

pocket ionization chambers sensitive to thermal neutrons. No permanent record of these 

dosimeter results was maintained, because the purpose of their use was to minimize exposure. 

Film badge results from this area for the first half of 1952 are provided in Table 6. 

Table 4: Exposure Analysis in the 86-inch Cyclotron Area for January 1, 1952–

July 1, 1952 (Struxness 1953) 

Exposure Range (mrem/week) Number of Exposures 

< 100 1,144 

100 – 300 88 

300 – 500 11 

500 – 1,000 4 

>1,000 3

In addition to external exposures, high airborne concentration levels existed in the alignment 

dock because of contaminated equipment held over during shutdown. During shutdown, air 

samples ranged from 16–2,077 dpm/m3. (Struxness 1953). Other radiological hazards, such as 

activation of equipment and materials, and potential exposures from target material have not 

sufficiently been discussed. 

Four additional accelerators used at Y-12 included the 63-inch Cyclotron, the 22-inch Cyclotron, 

the Cockcroft-Walton linear accelerator, and the Van De Graff accelerator. The 63-inch 

Cyclotron was designed to accelerate triply ionized nitrogen atoms at any energy up to 25 MeV 

at 1 to 10 microamperes. No shielding was planned for this unit (Struxness 1952), which was to 

begin operations in July 1952. The 22-inch Cyclotron was installed in one of the beta tracks for 

use as an aid in studying ion sources, beam deflection, and associated problems. This system

accelerated protons up to 2 MeV. The predominant radiation hazard was from x-ray generation. 

There were some issues with water leaks and liner damage associated with this unit (Livingston 

and Zucker 1961). A Cockcroft-Walton linear accelerator capable of producing a maximum of 

1 x 1010 fast neutrons per second and a 5-MeV Van de Graff accelerator capable of producing a 

fast neutron flux of 560 n/cm2

/s near the target, were also installed at the Y-12 facility (Kerr 

2005b). 

The main radiation hazards around these accelerators would be prompt gamma and neutron 

radiation, and decay gamma and beta radiation from activated targets and structural components. 

Additionally, operators and maintenance personnel that worked on the structural components and 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

82 of 203 

target holders may have been exposed to both external and internal radioactive compounds from

such things as radioactive copper and other metals during Dee repairs, electrical repairs, and 

other maintenance activities. This is particularly true of work inside the units. 

Incidents of internal exposures from working on such accelerators did occur at some other 

research facilities during this era. For example, at a 60-inch cyclotron in 1962, a crew of 10 

workers acquired body burdens of up to 9 nCi of 65Zn from inhaling fumes from soldering 

radioactive copper. Some accelerator workers at an 88-inch cyclotron showed positive for alpha,

beta, and gamma internally deposited radioisotopes (Patterson 1973, pgs. 528 and 529). 

The development of a D&D source of mono-energetic neutrons was mention in a 1950 letter 

from C.E. Center to S.R. Sapirie of the AEC (CCCD 1950) as a potential source of slow and fast 

neutrons. No other references to these experiments were found. 

5.6.3 Critical Assemblies 

A letter from the Pile Hazard Committee dated January 9, 1947 (Leverett 1947) discussed the 

safety issues with the pile that was being built at that time at Y-12. The Critical Experiments 

Facility (CEF) was operated at Y-12 by ORNL from 1950–1992. There were four major 

incidents at the facility during its operations, but no major personnel overexposures occurred as 

per E. Dickens (Dickens 2003, pg. 32). 

Prompt thermal/fast neutrons and gamma-rays could have been present in and around the CEF, 

as well as activation products that presented gamma, beta, and alpha exposure potentials. From

the information available (Kerr 2005b, page 20), it appears that, in the later years, they were 

aware of the radiation fields present at the CEF. However, not much has been said concerning 

the early years of the facility or if any prototype assemblies were constructed in the very early 

years at Y-12 (1947–1950) before the CEF was finished. 

In view of the potential internal and external exposures to Y-12 workers directly operating or 

supporting operations of these units, it would seem appropriate to include some discussion of 

Y-12 CEF and prototype assemblies, and the potential exposure conditions associated with each 

unit. Discussions on effectiveness of personnel monitoring, such as external dosimeters and 

bioassay techniques to determine exposure to personnel, should also be included as a part of the 

discussion. A portion of this work has been completed in ORAUT-RPRT-0033 (Kerr 2005b); 

however, this information is not integrated into the TBD or referenced by the TBD. 

5.7 ISSUE 7: DESCRIPTION OF HISTORIC Y-12 OPERATIONS AND 

RADIOLOGICAL CONTROL PRACTICES 

The TBDs do not go into enough depth on the varying and changing nature of Y-12 operations 

and work environments to provide the dose reconstructors with the specific knowledge that is 

needed for specific group or individual dose reconstructions. It is important to adequately 

understand the potential hazards and dose potential for each operation performed. Changes in 

operations that occurred over varying periods and the changing mission of facilities and 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

83 of 203 

buildings can also impact the dose that individual workers or groups of workers might have 

received. The ability to reconstruct dose in these many unique and varying types of functions 

related to a worker’s dose potential in a specific job is dependent on a better-developed 

description of individual or group activities and how their operations were modified by different 

mission redirection. 

5.7.1 Operational Process Descriptions 

The TBD acknowledges that buildings at Y-12 underwent several changes to accommodate 

mission demands. These changes should be more fully developed in the TBDs. Sufficient 

details should be provided about each process and the radiological practices employed at that 

time, and the degree to which these practices were effective in monitoring and minimizing doses 

to workers. 

Thousands of nuclear weapons were produced between 1945 and 1990. As the primary uranium

foundry complex for the nuclear weapons program, there is little doubt that Y-12 had a 

significant role in the production of the vast preponderance of U.S. nuclear warheads. In fact, as 

observed by Y-12’s operating contractor, BWXT, “Every weapon in the nation’s stockpile has 

some components manufactured at the Y-12 Complex” (BWXT 2003). 

In the late 1940s, the Y-12 Plant began making uranium weapons parts, which up to that point, 

had been made at Los Alamos. In 1953, the Y-12 Plant began producing lithium-6 and 

“secondaries” for the newly developed thermonuclear weapons. In 1962, the AEC decided to 

consolidate uranium component production at the Y-12 Plant, and existing capabilities at Rocky 

Flats were transferred to Y-12. Beginning in 1982, Y-12 began to store the stockpile of heavy 

water produced by SRS, a potential source of tritium. 

From 1943 to the present, Y-12 Plant’s activities have encompassed the following: 

• Separation of 235U (Jessen 2005) ― 1943–1947 

• Manufacturing fission-fueled nuclear warheads (Cochran 2005) ― early 1950s to the 

mid-1960s

• Manufacturing nuclear weapons components, including primaries, and thermonuclear 

secondaries (canned subassemblies (CSA)) (Cochran 1987; DOE 2002a) ― mid-1950s to 

1992) 

• “Special projects,” such as reactor fuel production (Klein 1994), and processing of 233U 

(West 1962; Cox 1997) ― 1950s and 1960s 

• Disassembling fission-fueled nuclear warheads (Norris 2004) ― early 1950s to 1992 

• Disassembling thermonuclear weapons secondaries (Norris 2004; DOE 2002b) ―

mid-1950s to the present 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

84 of 203 

• Storing weapons components, various highly-enriched uranium compounds (DOE 1996), 

small nuclear reactors (Klein 1994), spent reactor fuel, sealed sources and other materials 

containing actinides (DOE 1994b; DOE 2005a), and other nuclear materials (DOE 1996) 

― 1940s to the present 

Since 1994, when enriched uranium metal production was shut down for criticality safety 

reasons, Y-12 has been undergoing refurbishment and upgrades of processing equipment to 

resume production. There have been tons of highly-enriched uranium held-up in deteriorated 

process equipment and in packages, awaiting stabilization, packaging, and safe, secure storage 

(DOE 1996). 

The TBD does not provide complete information on the manufacturing, assembly, disassembly, 

and storage of nuclear weapons and weapons components. Specifically, the TBD does not 

provide (1) information regarding important process changes, “special projects,” and radiological 

incidents (2) complete data on radionuclides of concern, and (3) information regarding problems 

related to radiological controls and other exposure risks, including those in the present time

frame. 

Y-12 was actively involved in the dismantling of warheads since the end of the Cold War 

(Norris 2004). A part of these activities included disassembly of secondaries and subsequent 

storage. Highly enriched uranium is also stored at the Y-12 plant upon completion of 

disassembly. As of 1996, there was a backlog of units awaiting processing, stabilization and 

storage at the facility (DOE 1996). The TBD for the Y-12 plant should provide a chronology of 

the nuclear warhead types manufactured and disassembled at the site similar to that provided by 

NIOSH for the Iowa Army Ammunition Plant (Leonwich 2005). 

Given the magnitude of production of the U.S. nuclear arsenal, the complexity of the Y-12 

operations, and the deteriorated state of facilities (many between 40 and 62 years in age), the 

TBD only provides cursory information regarding processes that are necessary for the dose 

reconstructor to understand the degree and extent of radiological hazards to workers. 

Y-12’s “9212 complex” serves as an example of the complexity of the operations of many of the 

operations at Y-12 that need to be more fully described in the TBD. The TBD contains several 

references to 9212 and acknowledges that, “9212 housed the largest chemical operations for EU 

purification recovery and chemical conversion as well as normal and DU machining operations”

(Jessen 2005), and “more than 100 operations and processes were performed in this complex” 

(Jessen 2005). However, the TBD does not provide much detail of specific processes in this 

large and important operation. 

The HEU assessment for Building 9212 provides important details regarding “special 

operations” not included in the TBD, including: 

• Packaging and shipment of uranium metal and oxide for research/medical 

reactor fuel fabrication,
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

85 of 203 

• Operation of an evaporator for the processing of low uranium content 

solutions such as mop water...; 

• Receipt and storage of HEU from Rocky Flats; 

• Receipt and storage of solutions from the plant laboratory. (DOE 1996)

In terms of incidents at the 9212 facility, the HEU assessment notes the following: 

A review of the Occurrence Reporting System indicated that since 1990, there 

have been 130 reportable Off-normal events (category III) and eight Unusual 

Occurrences (Category II) pertaining to the enriched uranium operations at the 

9212 Complex. The preponderance of the Off-normal events was uranium 

contamination incidents (75). The next largest group of events were releases 

from exhaust stacks (17). 

Prior to the utilization of the occurrence reporting system, the following significant events took 

place in Building 9212 (DOE 1996):

• Improper uranium transfer (1990)

• Exceeding criticality safety limit in a 55-gallon combustible drum (1989) 

• Uranium release from stack 42 (1987)

• Operator uranium exposure (1986)

• Uranium chip fire (1985) 

• Uranium accumulation in exhaust stack (1983)

• Large geometry plastic bag used to collect uranyl nitrate leak (1967)

• Criticality Accident in C-1 wing (1958)

Significant information pertaining to specific activities and changes in Y-12’s numerous 

buildings that span the history of the site is missing. NIOSH should provide a greater depth of 

analysis on how these processes impact the dose to the worker. Without a better understanding 

of these details, the dose reconstructor is limited in his or her ability to adequately characterize 

dose to the claimant for all the different types of jobs performed and for potential exposure to 

situations that may not have been adequately monitored. 

5.7.2 Radiological Controls 

NIOSH has not provided enough information regarding the radiological controls in place (or lack 

thereof) for the operations at Y-12 that pose the greatest potential for significant exposures by 

job category and facility location. Especially in the early days, lack of adequate radiological 

controls can lead to worker doses that may be missed in the process of dose reconstruction. Such 

worker doses may not be identifiable without this understanding of the effectiveness of the 

radiological controls in place in specific locations, during specific evolutions, and for work 

practices where significant dose is possible. This appears to be even less developed for “legacy 

contamination.” 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

86 of 203 

Perhaps the most telling of a number of independent health physics reviews that have been 

conducted of Y-12 radiation protection practices was an April 1993 review that was conducted 

by professional staff of the DNFSB (1993). That these findings were made after the 

contamination control program upgrades of the mid to late-1980s are significant and indicative 

of the deep-rooted radiological management practices at the facility. This review found the 

following: 

...there are numerous instances where Y-12 is not in compliance with DOE Order 

5480.11, the DOE Radiological Controls Manual, or consensus standards. In 

many of these cases there is no documented technical justification for the

condition and no compensatory measures are in place. (DNFSB 1993) 

The review further noted the following: 

...it appears that DOE-ORO and MMES [Y-12 operating contractor] do not feel 

they have the resources to implement many of the mandatory requirements of the 

DOE Radiological Control Manual. (DNFSB 1993) 

The DNFSB staff review detailed a spectrum of long-standing programmatic deficiencies in the 

occupational radiation protection program:

• Contamination control: “Some of the most basic tenets of contamination control are 

not being followed in many facilities at Y-12. This condition greatly increases the 

possibility of the spread of contamination to uncontrolled areas and the unnecessary 

exposure of personnel to contamination” (DNFSB 1993). 

• Personnel monitoring: “Personnel monitoring is not currently required at egress points 

from all routinely occupied loose surface contamination areas at Y-12...” (DNFSB 1993) 

• Anti-contamination clothing: “Personnel in some known loose surface contamination 

areas at Y-12 (including areas where there is no egress monitoring) normally wear only 

lab coats and shoe covers as anti-contamination clothing” (DNFSB 1996). 

• Break areas: “...personnel are allowed to exit posted contamination areas and enter 

eating areas after surveying only their hands and donning a clean lab coat and shoe covers 

over potentially contaminated clothing” (DNFSB 1996). 

• Personnel decontamination: There are numerous boundary control stations at Y-12 

where personnel are required by procedure to wash their hands before monitoring for 

potential skin contamination [in violation of the DOE Radiological Controls Manual and 

standard industry practice]. 

• Training: Current training procedures and practices at Y-12 do not ensure radiation 

workers complete all radiation safety training required by DOE Order 5480.11 prior to 

being allowed unescorted access to radiological areas (DNFSB 1993). 

In 1995, the staff of the Energy Department’s Defense Nuclear Facility Safety Board reported 

the following: 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

87 of 203 

The mission most relevant to safety is one of processing the backlog of in-process 

materials at Y-12. In Building 9212, these materials occupy space in the 

hallways and operating corridors and some have been present for more than 

40 years. The in-process materials do not meet the criteria for interim or longterm storage and no criteria for in-process storage have been developed. Inprocess materials form the largest portion of the "material at risk" considered in 

Building 9212 accident analyses and contribute significantly to the dose 

consequences of those accidents. These materials pose the greatest risk for spills, 

decomposition, or criticality safety infractions, make inventories difficult, and 

increase worker exposure risks due to their location in the workplace. (DOE 

1995) [Emphasis added]

According to the DOE’s 1996 HEU vulnerability assessment, legacy contamination was 

considered a significant vulnerability to workers:

Beginning in 1988, significant upgrades to Radiological Control Programs were 

initiated. Prior to 1988 controls were in place for materials leaving the 

production area of the Y-12 plant. However, uranium contaminated items could 

be stored essentially anywhere within the western end of the plant. Uranium 

contaminated items could also be freely transported between production buildings 

without rigid controls to prevent the spread of uranium contamination. Because 

of these past practices, low-level alpha contamination exists in various places 

throughout the western end of the plant. The most logical spaces for uranium 

contamination to exist have been surveyed accordingly. 

However, a comprehensive survey has not yet been completed. Workers can be at 

risk from contamination that is the result of nearly 50 years of nuclear weapons 

manufacturing in two ways. First, there is always some risk of a small radiation 

exposure through inhalation or ingestion of uranium contamination outside 

radiological control areas. Secondly, the large operating spaces that are 

contaminated require in facility workers to be ‘fully dressed out’ in anti 

contamination clothing and sometimes respiratory protection for extended 

periods of time. In addition, the presence of radiological controls in outdoor 

areas (e.g., rooftops) present a potential for spread of contamination due to 

precipitation or windblown vegetation. [Emphasis added] (DOE 1996) 

Unfortunately, in recent times, unexpected exposure from uranium contamination has resulted in 

significant personnel exposures. The following describes a contamination incident that occurred 

in September 1998: 

...elevated levels of uranium were observed in bioassay data for one individual 

working in Building 9212 E-Wing casting operations...A total of 12 individuals 

are under radiological work restriction and about 60 operations and support 

personnel are being monitored for elevated levels. Possible causes under review 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

88 of 203 

include overall E-Wing contamination control and work practices in the casting 

knockout line and material handling areas in E-Wing” (DOE 1998a). 

Subsequently, the following was reported: 

...bioassay (fecal) results for one E-Wing worker have exceeded the LMES 

administrative control level of 1000 mrem/year (i.e., 1124 mrem CEDE). He and 

three other workers who are approaching this limit have had their administrative 

control level increased to 1500 mrem. (DOE 1998b) 

In March 2000, the inability to remove contamination sources persisted, as evidenced by the 

following: 

 In many of the unoccupied spaces around the Y-12 Plant, bulk waste containers 

and numerous bags have been accumulating (see item on 9212 E-Wing). Much of 

this waste is defective or obsolete hardware, renovation debris, or combustible 

trash, much of it potentially contaminated with uranium compounds. 

(DOE 2000a) 

Slow progress in nuclear material stabilization and removal has been exacerbated by difficulties 

in accessing historic process information. In 2002, the DNFSB staff noted the following (DOE 

2002b): 

On Monday, the site representative and an NNSA facility representative toured 

Building 9201-5 (Alpha-5). Of the approximately 500,000 ft2 structure, only 

about 10 - 20% is utilized by Depleted Uranium Operations to operate a vacuum 

arc remelt furnace and associated support operations. The remainder of the 

building contains material and equipment that was either abandoned in place or 

has been shifted from other facilities for indefinite storage. It is apparent that the 

very large inventory of deserted and abandoned material will be a challenge for 

future deactivation and decommissioning efforts. Of particular concern is the 

appropriate identification of hazards given the questionable control of previous 

shutdown activities. At the time parts of the facility were deactivated, the facility 

was parsed into approximately 80 “capability units” (CUs). A series of 

deactivation walkdown packages for ~25% of the CUs were completed by 1997, 

but funding limitations resulted in the termination of the effort. An additional 

source of information on potential hazards is from interviews of approximately 

240 veteran employees conducted under the Y-12 Knowledge Preservation 

Program during the 1993 to 1997 time frame. Transcripts of these interviews 

are searchable through a sophisticated database, but a priori knowledge is 

essential to understanding the content of the search results. It is apparent that 

information important for efficiently assessing safety of future 

decommissioning activities is dispersed among several organizations and will 

require a substantial effort to appropriately characterize hazards associated 

with decommissioning the facility. [Emphasis added] (DOE 2002b) 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

89 of 203 

The accumulation of HEU for storage at Y-12 has increased because of the return of weapons 

components, shipments of excess HEU from other DOE sites, and the safety-related delay in 

resuming uranium recovery operations. Despite the eventual full resumption of recovery 

operations in the 9212 building, HEU storage inventories may increase to the bounding quantity 

for off-site shipments of 98.4 metric tons (DOE 1996): 

Increased storage has the potential to increase the level of radioactivity exposure 

to workers, either due to higher volumes of material or storage of different types 

of material. The net effect is that the potential for higher levels of radioactivity 

exposure to workers could require new radiological controls or potentially result 

in higher exposure to workers. (DOE 1996) 

Unfortunately, the inability to remove and process potential sources of contamination continues 

to plague the site as of 2005. These long-standing problems can significantly impact the quality 

of dose reconstruction, which NIOSH should address. 

Building 9204-4 personnel determined that seven drums containing machine 

chips were not vented. Some of these unvented drums have been loaded since 

1990; several other drums containing depleted uranium chips are vented or have 

pressure relief devices. The chips are supposed to be in water but no drum 

inspections are known to have been performed and the water levels in the drums 

are not known. 

The chips are being stored as there is no current processing/disposition path for 

the chips due to a processing concern. A February 2001 Y-12 technical report 

was identified addressing depleted uranium machine chip storage. This report 

noted the pyrophoric and reactive (hydrogen generation in water) nature of the 

material. The report provided ‘recommendations’ that chips be stored in water 

with vented drums and that inspections be conducted on a regular basis to check 

for water level, drum condition and any chip corrosion or sludge. The report 

noted ‘Interim storage of chips under water is considered acceptable only when it 

is unavoidable.’ (DOE 2005b) 

The lack of radiological control, especially in the case of contamination control and 

environmental release, should be discussed in the TBD in relation to the potential for routine and 

special exposure conditions. These issues pose major obstacles to dose reconstruction by 

NIOSH. Of particular importance to dose reconstruction are the transcripts of interviews by 

veteran Y-12 employees generated by the “Y-12 Knowledge Preservation Program,” as noted 

above. NIOSH should seek to obtain these transcripts from the DOE in an expeditious manner. 

5.8 ISSUE 8: OCCUPATIONAL ENVIRONMENTAL DOSE 

This section of the Y-12 site profile review addresses the TBD for Y-12, Occupational 

Environmental Dose (Ijaz and Adler 2004). In the remainder of this section, the report of Ijaz 

and Adler (2004) will simply be referred to as the TBD. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

90 of 203 

5.8.1 Additional Exposure Pathways 

The TBD addresses only two exposure pathways; inhalation of the three naturally occurring 

uranium isotopes—234U, 235U, and 238

U—in ambient air, and external exposure to direct 

penetrating radiation. There are at least two other potential pathways that need to be addressed; 

(1) inadvertent ingestion of radioactively contaminated soil and other finely dispersed 

radioactive materials, and (2) inhalation of radionuclides other than the three naturally occurring 

uranium isotopes. These pathways are discussed below. 

5.8.1.1 Inadvertent Ingestion 

Whenever individuals perform outdoor work, there is a potential for ingestion of soil and other 

windblown particulates that are larger than the respirable particles. This pathway is traditionally 

included in dose assessments (see EPA 1997). This pathway is always difficult to assess, but a 

scoping calculation should be performed to place an upper bound on the dose that could be 

delivered. 

5.8.1.2 Other Radionuclides 

According to Ijaz and Adler (2004), Station HP-12 (also known as HP-32) was dedicated to 

monitoring for fission products. Since this station was just outside the Y-12 facility, and 

operated during the period 1959 – 1970, an effort should be made to uncover these data. 

Furthermore, Station 40, also on the Y-12 perimeter, monitored 131I, other fission products, and 

transuranics. The TBD should attempt to quantify or provide upper-bound estimates of the 

concentrations of such other nuclides. 

Buddenbaum et al. (1999) report that the X-10 facility handled irradiated thorium, radioactive 

lanthanum, and plutonium. Airborne effluent emissions from that plant included 233U. This 

isotope has a much higher specific activity, and somewhat higher effective dose coefficients, 

than the other uranium isotopes. Thus, if this nuclide was included in the total uranium,

expressed in :g/m3

, measured at the Y-12 facility, it would have a significantly greater 

radiological impact. Furthermore, when 233U is produced by irradiating 232

Th in a reactor, 232U is 

an inevitable by-product. Although its mass concentration is in the range of 5–50 ppm in the 

uranium metal, it has an extremely high specific activity, due to its short (68.9-y) half-life, and 

has effective dose coefficients that are several times higher than those of 234

U. 

Although the ORNL facility was about 6 km from Y-12, this source could prove to have a 

significant impact during periods of high releases, depending on the wind rose and other 

meteorological conditions. At the very least, the TBD should provide a scoping analysis to 

demonstrate that releases from facilities other than Y-12 would have no significant impact on 

workers on that site, if the authors have reason to believe that this is the case. 

5.8.2 Inhalation of Uranium 

The radiological assessment of the inhalation of uranium particulates on site at Y-12 involves 

two issues: (1) average activity concentrations of uranium isotopes in the ambient air, 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

91 of 203 

 

(2) chemical form (i.e., lung clearance type) and particle size. We will first discuss the 

methodology that the TBD used to calculate the historical uranium air concentrations. 

5.8.2.1 Average Activity Concentrations of Uranium 

The TBD used annual-average air concentrations of uranium, expressed or calculated as mass 

loadings (e.g., kg/m3

), from four stations within the Y-12 perimeter, coupled with estimates of 

the mass of uranium released each year from the Y-12 facility, to calculate values of annualaverage values of atmospheric dispersion coefficients (P/Qs) for uranium releases. Although this 

is a straightforward mathematical exercise, we question the validity of this empirical P/Q 

approach for estimating uranium concentrations in the ambient air for periods for which no air 

concentration data are reported in the TBD. 

Emprical P/Q Approach: General Observations

The empirical P/Q approach is based on several premises that are not explicitly stated and which 

need to be explored and justified before this approach could be accepted. 

(1) All the occupationally related airborne activity originates at Y-12. Y-12 is part of the 

ORR complex, which includes other sources of radioactive material. The varying 

correlation of the airborne uranium concentrations to releases from Y-12, especially at 

Station 12, calls this hypothesis into question. (See also the discussion under “Other 

Radionuclides” earlier in this section.) 

(2) In Section 4.2.5, the TBD attributes the poor correlation at Station 12 to the paucity of 

data. The smaller number of data points (8 vs a maximum of 17) does not account for the 

virtual lack of correlation shown in Figure 4.2.5-4. If anything, fewer data points would 

lead to a better correlation, both on mathematical grounds (fewer degrees of freedom), 

and on physical grounds, because meteorological conditions would vary less during a 

shorter time period. Despite this lack of correlation, the TBD uses the Station 12 data in 

creating a framework for dose reconstruction. 

(3) Resuspension of uranium deposited on soil, both inside the Y-12 perimeter and off site, is 

ignored. Resuspension of deposition during previous years could contribute to the 

uranium concentration in air, yet is uncorrelated to uranium releases from Y-12. 

Although one of the authors observed that the topography and ground cover of the area 

would lower the potential for resuspension,5

 scoping calculations or upper-bound 

estimates of the potential exposure from resuspension should be included in the TBD 

(Ijaz and Adler 2004). 

(4) The influence of other sources of uranium—releases from sources outside Y-12 or 

resuspension from onsite soils—is clearly indicated by the data for 1993. The uranium 

release of 3 kg for that year is the lowest during the period that Stations 2 and 12 were 

operating, and the second lowest during the 17-year operating period of Stations 4 and 8. 

The 1993 calculated P/Q's for all four stations are the highest of any year that these 

5

 See minutes of telephone conference among ORAU, NIOSH, and SC&A staff members dated June 10, 2005. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

92 of 203 

 

stations operated. The data from Station 2, which SC&A subjected to an independent 

statistical analysis, show a negative correlation between the calculated P/Q's and the 

annual releases of uranium―the lower the releases, the higher the P/Q. 

(5) The TBD states that the calculated P/Q's follow a lognormal distribution with a high-end 

tail. Yet, in assigning 95th percentile values, where the high-end tail would have the 

greatest influence, that aberration from lognormal behavior is ignored. An examination 

of the data in Tables 4.2.3-1/4 reveals the following: 

A. Station 2: Out of 12 calculated P/Q's, 1 is significantly higher than the 95th percentile 

value (2.00 H 10-14

 vs. 1.31 H 10-14). Statistically, the highest of 12 values would 

correspond to the 92nd percentile. 

B. Station 4: Out of 17 calculated P/Q's, 1 is over 1.7 times higher than the 95th

percentile value (8.73 H 10-14

 vs. 5.03 H 10-14). Statistically, the highest of 17 values 

would correspond to the 94th percentile. 

C. Station 8: Out of 17 calculated P/Q's, 2 are higher than the 95th percentile value. 

Statistically, the second highest of 17 values would correspond to the 88th percentile.

D. Station 12: Out of only 8 calculated P/Q's, 1 is higher than the 95th percentile value. 

Statistically, the highest of 8 values would correspond to the 88th percentile. 

(6) The above observations clearly indicate that the distributions are not lognormal, and that 

the assumptions used by ORAU to calculate the 95th percentile P/Q's are not claimant 

favorable, aside from the questions raised earlier in this discussion about the validity of 

the empirical P/Q approach. This conclusion is further buttressed by our statistical 

analysis of the Station 2 data, which exhibit the best correlation between the annual 

releases and the annual average air concentrations of uranium. Subjecting the calculated 

P/Q's to the W test (Gilbert 1987, Section 12.3.1) shows that there is a 10% to 50% 

probability that the distribution is lognormal. In other words, there is a better than a 50% 

chance that the distribution is not lognormal. 

(7) The estimates of uranium releases from Y-12 cited in Tables 4.2.3-1/-4 are taken from 

Buddenbaum et al. 1999. The estimates for the years 1983–1988 were independently 

calculated by these authors, and are, in general, significantly higher than the estimates 

provided by DOE. The releases for 1988 and later years are based on DOE reports and 

were not recalculated. The use of such mixed data in calculating the empirical P/Qs 

appears to make a significant contribution to the scatter in the results. The P/Qs for 

1983–1988 for the three stations—2, 4, and 8—for which such data are tabulated are 

listed in Table 5. The values for each station display coefficients of variation6

 with a 

range of 19%–34%, and an average of 25%. In contrast, the coefficients of variation for 

the years 1989–1999 (1989–1994 for Station 8) display coefficients of variation with a 

range of 136%–230%, and an average of 173%. 

6

 Defined as the ratio of the standard deviation to the mean
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

93 of 203 

Table 5: Calculated P/Q for Three Stations at Y-12 Facility (m-3) 

Station 

Year 2 4 8 Average

1983 4.37E-15 6.27E-15 1.24E-14

1984 4.98E-15 7.74E-15 7.19E-15

1985 5.08E-15 1.22E-14 1.24E-14

1986 4.75E-15 1.49E-14 1.11E-14

1987 2.68E-15 7.17E-15 8.08E-15

1988 3.94E-15 7.11E-15 8.38E-15

Mean 4.30E-15 9.23E-15 9.93E-15

Cv

a 19.1% 34.5% 21.3% 25.0%

1989 7.56E-15 1.89E-14 2.33E-14

1990 8.67E-15 1.29E-14 1.51E-14

1991 3.12E-15 4.44E-15 1.08E-14

1992 4.72E-15 2.06E-14 1.21E-14

1993 2.00E-14 5.03E-13 1.30E-13

1994 2.08E-15 5.00E-15 3.33E-15

1995 1.00E-13 4.00E-14

1996 9.00E-15 9.00E-15

1997 1.51E-14 2.50E-14

1998 1.10E-15 1.10E-14

1999 2.00E-14 2.00E-14

Mean 1.74E-14 6.09E-14 3.24E-14

Cv 155% 230% 136% 173%

Ratiob 4.0 6.6 3.3

a

 Coefficient of variation

b

 Ratio of 1983–1987 mean to 1988–1999 mean (1988–1994 for Station 8)

Clearly, the data for 1983–1988 show much better correlations between the measured 

concentrations and the estimated releases. For Stations 4 and 8, for which complete data are 

listed for 1983–1999, the ratios of the average P/Qs for 1989–1999 to the average P/Qs for 

1983–1988 are 4.0 and 6.6, respectively. There are three possible explanations for these results: 

• The releases for the years 1989–1999 have been under-reported 

• There were sources of uranium other than the releases from Y-12 for those years―such 

sources would make a greater contribution during the years following 1988, when the 

reported releases from Y-12 decreased dramatically 

• The method of measuring and/or calculating the airborne uranium concentrations changed 

during this period 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

94 of 203 

 

Utilization of All Available Data

The TBD states that uranium air concentrations were not available after 1999. However, the 

Tennessee Department of Environment and Conservation (TDEC) continued to collect samples 

and analyze them for gross " and gross $. Since the objective is to estimate doses from the 

inhalation of airborne activities, a conservative upper-bound estimate could be derived by 

assuming that all the gross " activity was 234U. Since the effective dose coefficients for the 

inhalation of this isotope are approximately 10% higher than those for 235U or 238U, such an 

approach would result in reasonable, yet claimant-favorable, estimated doses. These data should 

be listed in the TBD for use by dose reconstructors. 

Calculation of Uranium Concentrations in Air

As a general observation, we disagree with the decision to base the dose reconstruction on the 

calculated P/Q's, especially for years for which measurements of air concentrations exist. The 

purpose of dose reconstruction is to determine the doses to individual claimants, not to maintain 

a uniformity of approach, which disregards real data in favor of a statistical construct that might 

not be claimant favorable. 

The methodology for calculating the air concentrations of the uranium isotopes, which are listed 

in Attachment D to the TBD (Ijaz and Adler 2004), is unclear. Our attempts to replicate some of 

the data in Attachment D using the information in the main body of the TBD were unsuccessful. 

We have requested a more detailed explanation of this methodology,7

 but have not yet received a 

response. Furthermore, some of the data in Table 4.2.4-1, which are the basis of the data in 

Attachment D, are themselves suspect. Notably, based on the masses of the 238U releases listed 

in the fifth column of that table, the 238

U activities released in 1994 and 1995 should be 0.008 

and 6.7 H 10-4 Ci, respectively, rather than 0.002 and 0.0021 Ci, as listed. 

SC&A is puzzled by the decision of the authors to list combined activity concentrations for 234U 

and 235U. These data are based on the releases calculated by Buddenbaum et al. (1999). 

According to these authors: 

Uranium activity amounts for 234U and 235U were combined to add a level of 

conservatism to the Task 6 screening assessment. . . . It would be appropriate to 

evaluate these uranium isotopes separately during a refined dose assessment such 

as one associated with a complete dose reconstruction study. 

Listing these isotopes separately would result in a more comprehensive guide to dose 

reconstructors and would be consistent with profiles of other DOE sites. The data needed for 

such a listing are tabulated by Buddenbaum et al. (1999). 

We also question the statement, “...it is conservative to assume that the air concentrations 

reported from 1996 to 2002 are equal to the concentrations reported for 1995” (Ijaz and Adler 

7

 See minutes of telephone conference among ORAU, NIOSH, and SC&A staff members, June 10, 2005. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

95 of 203 

2004, pg. 26). First, there are no air concentration data listed for the years 2000–2002. The 

TBD states that the data for these years, collected by TDEC, are not usable. (We question that 

assessment, as discussed earlier in the present report). More important, this statement is 

contradicted by the data that are listed. The annual average concentrations at Station 4 for the 

years 1996–1998 are higher than for 1995. Furthermore, the data in Table 4.2.4-1 shows that the 

uranium releases in each year from 1996 through 2001 were higher than in 1995. Such 

discrepancies in the TBD are troubling, in that they may be indicative of a lack of technical 

review and/or quality control for at least this portion of the TBD. 

There is little guidance to dose reconstructors for applying the data tabulated in Attachment D. 

The tables list both 50th and 95th percentile values. Presumably, the 50th percentiles are to be 

used as the best estimates, and the 95th percentiles as the upper bound values. As mentioned 

earlier, the calculated 95th percentile values are not upper bounds, as they are exceeded by annual 

average P/Q's at each of the four monitoring stations. 

Finally, the contributions of 236U are ignored in Attachment D. Concentrations of 236U, which 

are listed in Tables 4.2.3-1/4, are comparable to those of 238U and should be taken into account.

Overall, the use of empirical P/Q's to estimate uranium concentrations in the ambient air does not 

provide scientifically valid or claimant-favorable guidance to dose reconstructors. 

5.8.2.2 Data Adequacy 

No data are furnished regarding the AMAD or the chemical form of the airborne uranium. As a 

general practice, NIOSH assumes an AMAD of 5 :m. However, Buddenbaum et al. (1999) 

states the following: 

Studies have been conducted at Y-12 to characterize uranium particle sizes in 

effluents. These studies indicated that, under normal conditions, uranium oxide 

particles were predominantly composed of small particles with typical mean 

diameters of 0.05 to 5 micrometers (millionths of a meter, :m). . . . Based on 

review of this information, one micrometer diameter uranium oxide particulates 

was the form of uranium assumed to have been released for the purposes of the 

Task 6 screening assessment that is presented in Section 4 of this report. 

Such information is required by dose reconstructors, who need to assign dose coefficients to the 

uranium isotopes. 

Finally, detailed information on the chemical forms of uranium should be furnished in the TBD 

(Ijaz and Adler 2004). NIOSH typically has assumed either Type M or Type S for inhaled 

uranium, using the form that is more claimant-favorable for a particular cancer. However, 

uranium in hexavalent form, including such compounds as UF6 and UO2F2, corresponds to Type 

F (ICRP 1994). Bruce et al. (1993) states the following: 

On May 11, 1956, a hydrogen line to the Building 9206 UF6 reduction area 

broke. This caused incomplete conversion of the UF6 to UF4. The chemical traps 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

96 of 203 

which were provided to capture the UF6 in such an emergency were filled to 

capacity at the time of the break, and UF6 was allowed to escape to the 

atmosphere. The escaping UF6 was visible as UO2F2 smoke as it was emitted 

from the vent stack... Other accidental releases of UF6 may have occurred at 

Y-12, but none were identified during this investigation. 

In the absence of more specific information, dose reconstructors should be instructed to consider 

all three lung clearance Types (F, M, and S) in selecting the most claimant-favorable effective 

dose coefficient to use in individual cases. 

5.8.3 External Exposure 

5.8.3.1 Utilization of All Available Data 

We question the conclusion in the TBD that the aerial surveys performed by the EG&G Remote 

Sensing Laboratory cannot be used for dose reconstruction. By spanning a period from 1973 to 

1989, they provide a historical record of the external exposures on the site. Such data could be 

used to provide a temporal trend, and should be combined with the 1985–1987 scoping survey. 

The spatial resolution of the aerial surveys is comparable to the size of the grid blocks in the 

scoping survey. Furthermore, the exposure rates calculated at 1 m above ground, as presented in 

the aerial survey reports, are more relevant to dose reconstruction than the ground level rates 

measured in the scoping survey. 

5.8.3.2 Calculation of External Dose 

We have questions about the methods used to convert the results of the scoping survey into dose 

rates. Given the purpose of the survey and the instruments used, we believe that the exposure 

rates were based on the count rates registered by the sodium iodide crystal in the survey meter. 

Such meters are commonly calibrated with a 137

Cs source—the scale on the rate meter is set to 

the calculated exposure rate at the location of the detector. This calibration is thus valid only for 

the principal 137Cs (-ray and will vary significantly with photons of different energies. The 

meter readings should be converted to dose rates in air (i.e., air kerma) using calibration curves, 

which can be obtained from the manufacturer of the meter in question or from manufacturers of 

similar instruments, and the known spectrum of the uranium isotopes deposited on the ground. 

(See the response curve for a 1” x 1” NaI crystal provided in Figure 9). The aerial survey data, 

which includes some spectral analysis, would be useful in this regard. The air kerma, in turn, 

should be converted to dose equivalent rates, using tables in ICRP Publication 74 (ICRP 1996) 

for the appropriate exposure geometry. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

97 of 203 

Figure 9: Energy Response of a Typical Survey Meter with a 1-inch NaI Crystal 

(Ludlum 2000) 

5.9 ISSUE 9: FREQUENT “INCIDENTAL” SOURCES OF CHRONIC 

WORKPLACE RADIATION EXPOSURE 

While we are aware that accidents and incidents are not addressed in site profiles, there is a class 

of incidents at Y-12 that were frequent enough to stand as a chronic source of acute workplace 

exposure, and widespread enough to affect a large number of workers in the work area involved. 

These include uranium and thorium chip fires, which were due to the pyrophoric nature of 

metallic uranium and the fine machining that was done with the material at Y-12. 

These fires, according to workers interviewed, occurred as often as several times an 8-hour shift, 

and ranged from small ones that the workers were responsible for extinguishing to larger ones 

that filled the entire operating area with uranium fumes. In some cases, the fires were intense 

enough to consume the lathe or machining tool itself. 

While special monitoring was not performed for workers exposed to chip fire air contaminants, 

air sampling was performed for major chip fires. The Y-12 Quarterly Health Physics report for 

the 4th quarter of 1962 (Y-12 1963) provides the following air sampling data taken during and 

after uranium chip fires in Building 9212. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

98 of 203 

Table 6: Air Sample Data During and After Uranium Chip Fire 

Location Samples % PAL d/m/m3

9212 Room 1008 (after chip fire and during cleanup) 4 100% 3,847 

9206 – Room 15 (during chip fire) 2 50% 140 

9206 – Room 100 (during chip fire) 1 100% 221 

While wearing of respirators and proper ventilation were recommended during cleanup, 

no records were found to confirm if precautions were taken by workers. 

Another example of a chronic source of “incidental” exposure of Y-12 workers are elevated 

airborne levels of uranium contamination due to failures of the building exhaust fans or, in the 

case of Building 9206, problems associated with the incinerator. The average quarterly airborne 

uranium activity (1962, 4th quarter) for the Mechanical Operations area of Building 9206 was 41 

d/m/M3

, which was almost 60% of the Y-12 airborne contamination PAL. The report concludes 

that most of this elevated contamination was due to “backflow of air through the ducts at a time

when the exhaust fans were off” (McClendon 1963). Given that the SC&A review of quarterly 

health physics reports was limited to a sampling of what was made available by Y-12, it is not 

clear what the historic frequency of these exhaust failures may have been; however, it is clear 

that they were treated as incidental occurrences despite their obvious contribution to worker 

exposure. 

Similarly, the Building 9206 incinerator was an intermittent source of airborne contamination in 

that building during its operation up until the mid-1980s. Various quarterly health physics 

reports from the early 1960s until the 1980s attribute elevated levels of airborne contamination to 

in-plant releases of uranium particulates from the incinerator in Room 25. In 1962, it was noted 

that the PAL for airborne uranium particulates (70 d/m/M3

) in Building 9206 had been exceeded 

for the 4th quarter, with an average reading of 150 d/m/M3

, with the “chief contributor” being the 

incinerator (the average airborne level in the incinerator room itself was 934 d/m/M3

 for the 

quarter). In 1985, it was noted that the airborne uranium contaminant levels would have 

exceeded Environmental Protection Agency then-proposed limit of 36 d/m/M3 for 7 months, 3 

quarters, and for the entire year of 1984 (West 1985). It was noted that a new incinerator had 

been scheduled for installation in FY1986. 

Another source of exposure is not addressed in the TBDs, but likewise could have lead to acute 

sources of internal exposure for workers involved with uranium and thorium melting and casting 

operations. Because the progeny of both uranium and thorium (radon and 226

Ra for the former, 

and thoron and 224Ra and 228Ra for the latter) have lower vaporization or melting points than their 

parent source materials, they are preferentially “boiled” off during arc furnace melting. For 

thorium, airborne activities of these daughter radionuclides have been measured as high as 

20,000 times the Maximum Permissible Limit as thorium metal ingots were removed from

casting molds (Ammann 1960). It is not clear from the TBD whether the potential significance 

of this exposure was addressed in the Y-12 external (for beta) and internal (particularly for radon 

and thoron) dosimetry programs. One of the Y-12 workers interviewed indicated that in-plant 

radon monitoring had been conducted, but that any positive readings had been attributed to 

natural sources, such as building materials. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

99 of 203 

The TBDs need to better characterize “incidental” sources of chronic workplace exposure. 

Situations that may not have been adequately covered by normal monitoring procedures need to 

be considered on an individual basis to assure correct dose assignment. 

Y-12 historically had no formal process for recording incidents prior to Occurrence Reporting. 

In some cases, the Radiological Control Organization generated reports or memorandums if the 

incident was considered “significant” (see Attachment 6). Although individuals may have been 

on monitoring programs during these incidents, it is important to ascertain any special exposure 

conditions which may have affected the accuracy of personnel monitoring such as nonuniform

exposure or geometry factor for external dose, or radionuclides, mode of intake, and chemical 

form for internal dose. Unmonitored or inadequately monitored workers may have been 

involved in undocumented incidents that may have resulted in unexpected exposures. Further 

consideration should be given to evaluating the availability of incident and work restriction data 

and its applicability to the dose reconstruction. 

5.10 ISSUE 10: SKIN AND EXTREMITY DOSES 

The TBD does not provide a technical basis for excluding non-penetrating doses from acute nonroutine exposures. Starting in 1948, the film badges used at Y-12 had an open window and a 

cadmium filter to aid in distinguishing between beta and photon exposures (Kerr 2003, pg. 13). 

With guidance provided in Attachment F of the TBD, it is unclear why non-penetration dose is 

not assigned, given the availability of data. There is also no information available on shallow 

dose exposures to skin of the extremity; although the original extremity dosimeter was used as 

early as the 1940s for some workers. 

The Y-12 Plant handled a variety of radionuclides over the period of operations. Many of these 

radionuclides or their progeny were beta emitters. Depleted uranium and enriched uranium have 

been the predominant materials processed at Y-12. These radionuclides can result in beta dose to 

the skin, and in some cases, the breast and testicles. The beta emitters of interest from depleted 

uranium are 234Th and 234mPa resulting from the decay of 238U. For 235U, 231

Th is the primary 

source of beta exposure. Although enriched uranium results in some beta exposure, depleted 

uranium is the primary beta hazard, due to the high-energy beta and decay rate of 234mPa. Dose 

rates to the extremities from handling can be considerably higher due to the proximity of the 

extremity verses the whole body (Thomas and Bogard 1994). 

Processing and handling of thorium and 233U also lead to beta exposure. The 232U impurity in 

233U presents both a beta and gamma exposure hazard. As the 232U concentration increases, the 

dose rates associated with the material increase. The 232U daughters 212Bi and 208

Tl emit 

energetic betas with maximum energies of 2.26 MeV and 1.8 MeV, respectively. For example, 

the beta dose rate at 1 foot from a 1-kilogram thin disk of 233U with 1 part per million 232U is 

10.3 mrad/hour. The total surface dose rate is 1.2 rad/hour for the same disk aged for 60 days 

(Owen 1964). Thorium-232 and daughters emit betas with maximum energies of 12 to 

2,260 keV. Calculated surface dose rates from 232Th in equilibrium with its daughters have been 

reported as 107 mrads per hour. External exposure potential can increase with the concentration 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

100 of 203 

or separation of thorium daughters (West 1965). Closer proximity to these materials yields 

higher beta exposure, such as with extremities during handling operations. 

5.10.1 Acute Beta Doses and Routine Nonpenetrating Doses 

The TBD appears to indicate unspecified issues with the estimation of non-penetration doses. 

Those performing waist-level uranium-handling jobs could have received beta doses (both acute 

and chronic) to the skin, testicles, and breasts from their work. Also, beta dose to the extremities 

may have been considerably larger than whole-body non-penetrating dose, especially for workers 

handling radioactive material. The TBD (Kerr 2003, pg. 9) states the following: 

Information presented in this section pertains to the analysis of these records and 

does not address parameters regarding skin and testicular or breast radiation 

dose that may result from acute exposure to beta-particles in generally nonroutine workplace exposure situations. 

Attachment F of the TBD (Kerr 2003, pg. 41) gives the following directions with respect to 

estimation of organ dose in cases of skin, testicular, and breast cancer: 

The information needed to evaluate claims is directed to the technical parameters 

of the annual estimates of the primary organ dose that is calculated from the 

dosimeter interpreted personal dose equivalent, Hp(10), and Hp(0.07) in the case 

of skin, testicular, and breast cancer. These are used as a consistent basis of 

comparison for all years of Y-12 occupational external dose starting in 1950. 

It is unclear from the passages above whether this means the record for non-penetrating dose 

should be considered adequate or inadequate. SC&A notes that Kerr (2003) does not direct the 

dose reconstructor to Addendum F of ORAUT-PROC-0006, Dose Reconstruction Project for 

NIOSH – External Dose Reconstruction (Merwin 2003), which provides a procedure for shallow 

dose calculations as follows: 

Under certain limitations, it has been determined that the calculated doses for 

skin cancers can be assured of being claimant-favorable if a factor of 2 is applied 

to deep doses reported by the site post-1980 (and, for 1970 and later cases if an 

organ DCF is also applied), and when shallow doses are calculated according to 

the instructions provided below. Therefore, external doses of complex-wide skin, 

breast, and testicular cancers can be calculated according to this procedure. 

OCAS-IG-001, External Dose Reconstruction Implementation Guide (NIOSH 2002), emphasizes 

that the external electron exposures are to be considered in the case of skin cancer. In the cases 

of breast and testicular cancers, electron exposure evaluations for beta particles with energies of 

> 1 MeV should be conducted. OCAS-IG-001 does not state that dose reconstruction for skin, 

testicular, or breast cancer is excluded for “acute exposure to beta-particles in generally nonroutine workplace exposure situations” (Kerr 2003, pg. 9). The TBD should explain why this 

exclusion has been made and provide the technical basis for the decision. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

101 of 203 

5.10.2 Extremity Exposure 

Both depleted uranium and thorium billet handling has associated high beta radiation fields 

leading to elevated hand and arm exposures for material handlers. The original extremity 

dosimeter used at Y-12 in the 1940s was crude and consisted of film attached to the hand with 

electrical tape. Later, some chemical operators began wearing dosimeter rings when handling 

transuranics, particularly plutonium. From film badge readings, Y-12 distinguished between 

“penetrating” dose and “skin” dose. Recorded average skin dose was highest for mechanical

operators (i.e., machinists) and product certification personnel. For five quarters in 1961–1962, 

the average skin verses penetrating dose by work category was as follows: 

Table 7: 1961–1962 Average Skin and Penetrating Dose by Worker Category 

Job Title Sample Size Skin Dose (mrem) Penetrating Dose (mrem) 

Chemical Operations 523 125 60 

Mechanical Operations 1,750 280 60 

Product Certification 451 330 60 

In terms of skin-penetrating radiation ratio, the machinists and product certification personnel 

had higher ratios of 4:1, as compared with chemical operators at close to 2:1. 

The source of the radiation is obvious. For depleted uranium billets in the Y-12 foundry, the 

surface readings for some were as high as 2 R/hour. Control actions were necessitated to provide

shielding for handlers and nearby workers, including covering the billets and setting up partial 

shielding (McLendon 1960). For thorium, there is an external exposure potential due to four 

beta emitters with maximum energies ranging from 12 to 2,260 keV. As detailed in Health 

Physics Considerations Associated with Thorium Processing, calculated beta dose rates at the 

surface of 232

Th metal in equilibrium with its daughters and 3 years after separation from its 

progeny “has been reported as 107 and 40 mrads [per hour], respectively” (West 1965). It is 

noted that this compares with 240 mrads from uranium in equilibrium with its daughters (West 

1960). 

The TBD should explain why extremity exposure data is excluded from consideration and 

provide the technical basis for the decision. Verbal communications from ORAU indicate that 

extremity dose considerations are being considered for inclusion in the next revision of the TBD. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

102 of 203 

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 with respect to 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, i.e., completeness of data sources, technical 

accuracy, adequacy of data, site profile consistency, and regulatory compliance. The SC&A 

review found that the NIOSH site profile (and its constituent TBDs) for Y-12 represents an 

adequate accounting of the “core” uranium exposure and dosimetric history of the plant, but falls 

short in fully characterizing underlying issues that are fundamental to guiding dose 

reconstruction. In some cases, these issues will impact other site profiles. Many of the issues 

involve lack of sufficient conservatism in key assumptions or estimation approaches, incomplete 

site data or incomplete analysis of that data, or incomplete reflection of operational or dosimetry 

history. Section 6.0 summarizes the key issues. Detailed evaluation of these issues is provided 

elsewhere in the report. 

6.1.1 Objective 1: Completeness of Data Sources 

The breadth of data sources used as a basis for the Y-12 site profile is evident in the 223 reports 

cited as references, including historical reports, correspondence, and other technical documents. 

The ORAU team includes health physics personnel with long histories within the Oak Ridge 

complex who have extensive knowledge of key dosimetry assessments and other documentation, 

which aided in identifying and gaining access to data sources. As noted in Section 4.0, NIOSH 

effectively compiled facility specific information for major facilities, and proceeded to 

characterize the types and relative importance of various radiological hazards that may have 

contributed to internal and external dose. Furthermore, NIOSH/ORAU have recognized gaps in 

the existing TBDs, and participated in follow-up data captures at the Y-12 and development of 

supporting technical documents. 

However, the site profile falls short in its critical evaluation of pertinent records and purposeful 

use of site expert interviews to ascertain potential monitoring or records gaps throughout Y-12’s 

lengthy history, with an objective of determining the extent and significance of missing or 

unmonitored worker dose. SC&A found that existing unclassified and classified information 

sources point to a number of significant sources of such gaps, including unmonitored support 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

103 of 203 

service workers, presence of radionuclides other than uranium, and unrecorded chronic low level 

uranium uptakes. These issues are directly pertinent to the conduct of dose reconstructions, and 

will require further records review and evaluation on the part of NIOSH (some of which has 

already been initiated with respect to internal dose assessments for thorium, 233U, and recycled 

uranium). 

Furthermore, the TBDs do not go into enough depth on the varying and changing nature of Y-12 

operations and work environments to provide the dose reconstructors with the specific 

knowledge that is needed for specific group or individual dose reconstructions. It is important to 

adequately understand the potential hazards and dose potential for each operation performed. 

Changes in operations that occurred over varying periods and the changing mission of facilities 

and buildings can also impact the dose that individual workers or groups of workers might have 

received. The ability to reconstruct dose in these many unique and varying types of functions 

related to a worker’s dose potential in a specific job is dependent on a better-developed 

description of individual or group activities and how their operations were modified by different 

mission redirection. 

Radioisotopes other than uranium requiring further assessment include 3

H, 90Sr, 99Tc, 210Po, 

228Th, 232

Th, 239Pu, 241Pu, 237Np, 233U, and 241Am. Some of these radionuclides were associated 

with research and development activities, while others were handled in production, either as a 

source material or as a contaminant, e.g., from recycled uranium. Given their potential radiation 

exposure significance, thorium, 233U, and transuranic handling is of particular importance to the 

TBD. While some of these operations are associated with ORNL missions at Y-12, a number of 

Y-12 workers (e.g., janitors, maintenance, and crafts personnel, as noted above) supported these 

operations and were potentially exposed to these sources. In addition, the environmental TBD 

has not considered release of radioactive materials other than uranium although processing 

involved other radionuclides. 

Although radiation generating devices, including x-ray-producing equipment and accelerators, 

were associated with operations at Y-12, there is little mention of these units, the radiological 

hazards associated with them, and the effectiveness of the dosimeters to measure these 

radiations. There is also no mention of the potential internal exposures from target material and 

maintenance activities. 

The TBD has included a discussion of the 1958 criticality accident involving overexposures to 

eight individuals. There are, however, other frequent “incidental” sources of workplace radiation 

exposure involving many workers over the course of operation that are not addressed in the site 

profile. These incidents involved uranium chip fires, failure of engineering controls, and 

vaporization of radium and its progeny during molten uranium and thorium casing operations. It 

is not clear whether the frequency and sensitivity of the historic Y-12 bioassay program would 

have detected these “spike” releases, particularly where urinalysis was delayed or unduly relied 

upon for Type S uranium oxides. Such releases, in particular, would need to be addressed where 

routine air sampling data is used to estimate potential worker uptakes. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

104 of 203 

6.1.2 Objective 2: Technical Accuracy 

The site profile accurately characterizes Y-12’s radiation dosimetric practices over its history, 

but fails to sufficiently scrutinize past assumptions and practices, which would be considered 

non-conservative in current accepted practice. As noted earlier, an overriding assumption that 

guided the early and mid-history of the plant was the prevailing acceptance of chronic low-level 

uranium uptakes as insignificant hazards and not subject, in practice, to routine ALARA 

considerations, exposure assessment, and dose estimation. With this historic context in mind, 

SC&A examined all historic assumptions that guided both internal and external dose assessment 

for uranium at the chronic, lower concentrations at Y-12. Interviews were conducted with 

longstanding workers, historic internal plant correspondence was reviewed, and site health 

physics personnel were interviewed regarding practices. It was found that the TBDs seemingly 

have the same “blind spot” that existed in the former Y-12 health physics program with respect 

to occupational low-level chronic uranium exposure. 

For example, the uncertainties in the bioassay techniques and detection limits used to quantify 

internal dose are significant issues in dose reconstruction and are not fully addressed in the TBD. 

The TBD intake model does not consider the 48-hour delay imposed on obtaining urine samples, 

which leads to an under-calculation of the chronic daily intake rate for uranium by a factor of 2–

4, depending on lung clearance class. It neglects the ingestion pathway, despite the fact that 

eating, drinking, and smoking were routine in contaminated operating areas of the plant until 

1988–1989. Dose assignments do not consider the measured particle sizes, the most claimantfavorable solubility type for the organ of concern, and the uncertainties associated with bioassay 

techniques and sampling methodology. 

While the TBD concludes that an appreciable percentage of neutrons were detectable at Y-12 

given the “hard” energies involved, this assumption is based on a limited amount of neutron 

spectral measurements conducted in 1990, and a more recent PNNL study at three specific 

locations. A full analysis of all potential neutron exposure conditions at Y-12, some elaborated 

on in ORAUT-RPRT-0033 (Kerr 2005b), has not been completed. The assumption that most 

neutron exposure at Y-12 was from neutron energies greater than 500 keV, and thus would be 

detected by the NTA film, is not adequately substantiated. There has been no consideration in 

the TBD for the poor detection of neutrons between 500 keV and 800 keV by film badges. 

The occupational environmental dose TBD (Ijaz and Adler 2004) leaves inadvertent ingestion of 

radioactively contaminated soil and inhalation of radionuclides other than uranium inadequately 

developed. NIOSH needs to examine more closely the radiological assessment of the inhalation 

of uranium particulates onsite at Y-12, which represents a source of worker exposure. The 

empirical χ/Q approach is based on several premises that are not explicitly stated, and which 

need to be explored and justified before this approach can be accepted. 

Extremity and skin doses should be given attention in the external dose TBD. Depleted uranium

metal handling has associated high beta radiation fields leading to elevated hand and arm

exposures of material handlers. Similarly, 232

Th has beta-emitting progeny that present a 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

105 of 203 

radiological hazard for direct handling. Since few workers wore ring dosimeters, it is important 

to characterize potential beta/gamma fields, and worker handling practice and geometry. 

6.1.3 Objective 3: Adequacy of Data 

Data gaps were evident in the TBDs, which were not adequately addressed in attendant analysis. 

For example, while NIOSH accepts that less than 25% of all Y-12 workers were monitored 

externally from 1950–1961, and no external data is readily available for 1948–1950, it assumes 

these individuals are the “maximally exposed” workers given the stated policy of the time. 

However, given that line supervisors made all badging decisions for groups and individuals 

within groups (Patterson 1957) in an era where production often took precedence over safety, 

legitimate doubts can be raised over how much weight should be afforded these management 

assignments. Historic discrepancies in recording badge readings also raise questions. For 

example, beginning in mid-1956, all film badges reading, “below the minimum detectable [were 

directed to be] recorded as the average of the minimum detectable reading and zero, instead of 

being recorded as the minimum detectable” (West 1956). In addition, “some fraction” of overall 

Y-12 film badge results were found to be flawed during 1950–1955 due to incorrect assignment 

of badges based on an incorrect determination of whether gamma or beta radiation exposure was 

predominant for a worker (West 1991). While documentation may exist to corroborate or correct 

for badging assignments, this information is not provided as a basis for applying the co-worker 

dose estimation model. 

Similarly, while production workers may have participated in routine monitoring programs, 

routine bioassay sampling among support workers and many crafts, who also had full access to 

radiological areas onsite, was not implemented prior to 1994. Questions arise as to the 

effectiveness of the bioassay methods to detect highly insoluble (“high-fired”) uranium oxides to 

which some workers were exposed either routinely or during incidents. This is particularly the 

case prior to the routine use of fecal and/or lung counting, as urinalysis can be insensitive to 

highly insoluble uranium. Two-day delays in sampling and limited sampling frequencies may 

not have been adequate to detect acute intakes of Type F materials. 

6.2 OBJECTIVE 4: CONSISTENCY AMONG SITE PROFILES 

Y-12, Hanford, and the SRS are all DOE facilities that have handled uranium, thorium, tritium,

and transuranics. The 300 Area of both Hanford and SRS handled large amounts of uranium,

which is often the predominant radionuclide in these areas. Y-12 primarily processed uranium;

however, there were operations involving thorium, tritium, and plutonium. Although the 

operations are not identical at the three sites, the general approach to monitoring at the three sites 

was similar. In addition, considerable material received by Y-12 was shipped to and from the 

SRS. NIOSH has appreciated the distinctions between each of the three sites, and tailored its 

TBD assumptions and analytic approaches to the unique histories and conditions at each site, 

while mirroring those assumptions and approaches where justified. The SRS and Hanford site 

profiles predate the Y-12 site profile; therefore, NIOSH benefited greatly from these earlier 

efforts. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

106 of 203 

An extensive comparison was performed by SC&A to compare and contrast the methodologies 

used in the Y-12, Hanford, and SRS TBDs to determine external, internal, medical, and 

environmental dose. These comparisons focus on the methodologies and assumptions associated 

with dose assessments and the derivation of values used to obtain a probability of causation for 

individual claimants. A detailed analysis is provided in Attachment 7 to this report. This table 

demonstrates, in detail, where the Y-12, Hanford, and SRS site profiles differ or agree on a 

number of important assumptions. 

There are some inconsistencies noted between the Y-12 site profile and other site profiles. For 

example, the Hanford and SRS site profile discount the use of NTA film as an adequate measure 

of neutron dose in favor of applying neutron-to-photon ratios. The Y-12 site profile, in contrast, 

defends the use of NTA film based on a limited neutron spectral measurement, and applies a 

correction factor to account for the undetectable lower-energy portion of the neutron spectrum. 

Likewise, the Y-12 TBD has not examined the film badge corrections for low-energy photons, 

such as those encountered with x-ray units. 

Hanford, SRS, and Y-12 all have Type F material, but the Y-12 TBD discounts the significance 

of this solubility type at Y-12. This position is questionable because the Type F solubility type is

more claimant-favorable for some organs. 

ORAUT-TKBS-0014-6 (Kerr 2003, pg. 31) states that missing doses can be calculated by 

multiplying the MDL by the number of zero dose results. In the Hanford document, ORAUTTKBS-0006-6 (Kerr 2003, pp. 47 and 75), and the Nevada Test Site document ORAUT-TKBS0008-6 (Rollins 2004, pg. 14), it is suggested that the missing doses be calculated by multiplying 

the MDL by the number of zero dose results divided by 2. This appears to be an inconsistency 

between these respective site TBDs that needs to be corrected or explained. 

6.2.1 Objective 5: Regulatory Compliance 

With one exception, NIOSH has complied with the hierarchy of data required under 42 CFR 

Part 82 and its implementation guides. 42 CFR Part 82 recommends the use of the default 5 μm 

particle size only in cases were there is no information on particle sizes. This is not the case with 

the Y-12 facility, as stated in ORAUT-TKBS-0014-5 (Rich and Chew 2005, page 10). 

6.3 USABILITY OF SITE PROFILE FOR INTENDED PURPOSE 

SC&A has identified seven criteria that reflect the intent of the EEOICPA, the Final Rule, 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 Y-12 site profile TBDs to determine whether they meet these criteria: 

Objective 1 − Determine the degree to which procedures support a process that is expeditious 

and timely for dose reconstruction. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

107 of 203 

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 DOL probability of causation 

estimate at the upper 99% confidence level. 

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. 

6.3.1 Ambiguous Dose Reconstruction Direction 

The external dosimetry TBD (Kerr 2003) lacks clarity and specific instructions regarding the 

assignment of external dose, specifically application of correction factors to recorded and missed 

dose. NIOSH has not provided the requisite instructions to ensure that workers who perform

waist-high operations are assigned an appropriate dose estimate as claimants who have 

developed skin, testicular, or breast cancer. There is a need for further development in the TBD 

regarding geometric correction factors and how they are derived from calibration and workplace 

experiences. Additional elaboration on the origin of the correction factors is needed to evaluate 

their validity. Furthermore, there is no clear guidance on when to apply the geometric correction 

factors. 

Table 6.3.3.2-1 (page 19 of the TBD) states that the angular response of Y-12 neutron 

dosimeters is likely too low, because of its lower response at non A-P angles (the text before this 

table discusses both NTA film and TLNDs). However, Table 6.3.4.3-1 (page 26) states that the 

NTA film response is likely too high, because of its increased response at other than A-P angles, 

and that the TLND response would decrease at other than A-P angles. This conflict needs to be 

resolved or clarified. 

The recorded neutron or missed neutron dose is to be adjusted by the ICRP 60 correction factors 

in Table F-5 (Kerr 2003, page 44) and an example for a worker in the storage area is provided on 

page 43. Additionally, ORAUT-TKBS-0014-6 (Kerr 2003, page 45) instructs the dose 

reconstructor to multiply the MDL by a factor of 1.05 or 1.10, depending on location, but does 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

108 of 203 

not include the ICRP 60 correction factors. It is unclear from reading the TBDs what correction 

factors are to be applied when, and if they are (or are not) to be applied sequentially. 

The internal dosimetry TBD (Rich and Chew 2005) provides a good description of the historic 

internal dosimetry program; however, it is unclear from the TBD what data is used in the 

evaluation of internal dose and how missed internal dose is assigned. NIOSH has clearly 

outlined the process to be applied for deriving co-worker doses where monitoring was lacking; 

however, the corresponding assumptions used to derive dose to monitored workers is unclear. 

6.3.2 Inconsistencies and Editorial Errors in the Site Profiles 

There are numerous editorial errors and inconsistencies in the TBD causing confusion. Although 

there are numerous TIBs or other technical reports that supplement the directions in the TBD, 

there is no reference to these. Attachment 9 provides a complete list of the editorial errors in the 

TBDs identified during this review. 

6.4 UNRESOLVED POLICY OR GENERIC TECHNICAL ISSUES 

A number of issues were identified that are common in the Y-12, Hanford, and SRS site profiles 

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), 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 

indicates that it may develop separate TIBs in order to address these more generic issues. The 

following represents those issues identified in the Y-12 site profile review that SC&A believes 

represent transcendent issues that need to be considered by NIOSH as unresolved policy or 

generic technical issues.

(1) Direction on the applicability of the TBD and/or TIBs to individual dose reconstructions 

is absent. 

(2) Mobility of 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. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

109 of 203 

(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 may be often 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 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. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

110 of 203 

7.0 REFERENCES 

42 CFR Part 82, Title 42, Public Health, Part 82, Methods for Radiation Dose Reconstruction 

Under the Energy Employees Occupational Illness Compensation Program Act of 2000, Code of 

Federal Regulations, May 2, 2002. 

Amman, P.R., Memorandum from P.R. Amman to L.E. Burkhart, Health Physics Aspects of 

Thorium Processing in Summary, Union Carbide Nuclear Division, Oak Ridge, Tennessee, 

March 31, 1960. 

Ashley, J.C., Barber, J.M., Snapp, L.M., Turner, J.E., 1992. Technical Basis for the Internal

Dosimetry Program at the Y-12 Plant. Y/DQ-40, Martin Marietta Energy Systems, Inc., Oak 

Ridge, Tennessee, March 2, 1992. 

Berger, C.D. and Lane, B.H., 1985. Neutron Dosimetry with the ORNL Badge, ORNL/TM9450, Oak Ridge National Laboratory, Oak Ridge, Tennessee. 

Bihl, D. E. 2004. Technical Basis Document for the Hanford Site – Occupational Internal Dose, 

ORAUT-TKBS-0006-5, Revision 01, November 14, 2004, Oak Ridge Associated Universities, 

Oak Ridge, Tennessee. 

Brackett, E. M. 2003. Technical Information Bulletin: Maximum Internal Dose Estimates for 

Savannah River (SRS) Claims, ORAUT-OTIB-0001, Oak Ridge Associated Universities, Oak 

Ridge, Tennessee. 

Brackett, E. M. 2005. Internal Dosimetry Coworker Data for Y-12, ORAUT-OTIB-0029, 

Revision 00, April 5, 2005, Oak Ridge Associated Universities, Oak Ridge, Tennessee. 

Bruce, G. M., et al. 1993. “Oak Ridge Health Studies: Phase I Report.” Vol. 2, Part A. Dose 

Reconstruction Feasibility Study— Tasks I & 2: A Summary of Historical Activities on the Oak 

Ridge Reservation with Emphasis on Information Concerning Off-Site Emissions of Hazardous 

Materials. <http://cedr.lbl.gov/DR/OAKPDF/historicalops.pdf> (June 13, 2005). 

Buddenbaum, J. E., et al. 1999. Task 6 Report: Uranium Releases from the Oak Ridge 

Reservation—A Review of the Quality of Historical Effluent Monitoring Data and a Screening 

Evaluation of Potential Off-Site Exposures. 

<http://www2.state.tn.us/health/CEDS/OakRidge/Uranium.pdf> (June 13, 2005). 

BWXT 2003. Y-12 National Security Complex, BWXT Y-12 Public and Government 

Affairs, U.S. Government Printing Office #2003-553-246,Oak Ridge, Tennessee. 

Center C.E.1950. Center C.E. to the Atomic Energy Commission, “The Control of Radiation 

Hazards,” May 5, 1950. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

111 of 203 

Center for Disease Control and Prevention (CDC) 2001. Savannah River Site Dose 

Reconstruction Project Phase II: Source Term Calculation and Ingestion Pathway Data 

Retrieval, Evaluation of Materials Released for the Savannah River Site, RAC Report No.1, 

CDC-SRS-1999, Center for Disease Control and Prevention, Atlanta, GA, April. 

Carbide and Carbon Chemicals Division (CCCD) 1950. C. E. Center Letter to S. R. Sapirie, at 

the Atomic Energy Commission forwarding a Report: Control of Radioactive Hazards by 

Carbide and Carbon Chemical Division, Carbide and Carbon Chemicals Corporation, May 11, 

1950. 

Carbide and Carbon Chemicals Corporation (CCCC) 1951. Process Report Health Physics;

November 1, 1950 – December 31, 1950,, Y-780, Carbide and Carbon Chemicals Company , 

Oak Ridge, Tennessee, July 9, 1951. 

Cochran, T., et al. 1987. Nuclear Weapons Data Book, Nuclear Warhead Production, Vol. II, 

Ballinger Press (1987) 

Cochran, T. 2005. Personal Communications with Thomas B. Cochran, Director, Nuclear 

Program Natural Resources Defense Council, June 15, 2005. 

Cofield, R. E., 1961. In Vivo Gamma Spectrometry for Inhalations of Neptunium-237-

Protactinium-230, Cobalt-60, and Zirconium-95-Niobium-95, Y-EH-679, Union Carbide 

Nuclear Company, Oak Ridge, Tennessee. 

Cox, S.O., Duerksen, W.K., Redmond, R.R., Vath, J.E., 1997. Criteria for Acceptance and 

Technical Assessment for Acceptance of Enriched Uranium at the Y-2 Plant, Y/LB-15-

15,920/Rev., March 1997. 

Cummins, C. L., C. S. Hetrick, and D. K. Martin, 1991. Radioactive Releases from the 

Savannah River Plant, 1954–1989, Environmental Protection Agency Summary, WSRC-RP-91-

684. 

Defense Nuclear Facilities Safety Board (DNFSB) 1993. Memorandum from J.J. McConnell, 

Oak Ridge Program Manager to G.W. Cunningham, Technical Director; Subject: Report to the 

Oak Ridge Y-12 Plant; April 9, 1993. 

Dickens, F., and Fleming, K. 2003. Technical Basis Document for the Oak Ridge National 

Laboratory – Site Description, ORAUT-TKBS-0012-2, Rev. 00, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee, November 14, 2003. 

Duncan, F. 2003. Technical Information Bulletin: Savannah River Site Tritium Dose

Assignment, ORAUT-OTIB-0003, Oak Ridge Associated Universities, Oak Ridge, Tennessee. 

Eckerman, K. F. and G. D. Kerr, 1999. Y-12 Uranium Exposure Study, “ORNL/TM-1999/114, 

Oak Ridge National Laboratory, Oak Ridge, Tennessee, August 5, 1999 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

112 of 203 

Emerson, L. C., 1951. Health Physics Report; November 1 – December 31, 1950, Union 

Carbide Corporation, Oak Ridge, Tennessee, July 11, 1951. 

Environmental Protection Agency (U.S.) (EPA). 1997. Exposure Factors Handbook. 

EPA/600/P-95/002Fa. Washington, DC: Author. 

Fix, J. J., E. S. Gilbert, and W. V. Baumgartner, 1994. An Assessment of Bias and Uncertainty in 

Recorded Dose from External Sources of Radiation for Workers at the Hanford Site, PNNL11196, Pacific Northwest National Laboratory, Richland, Washington. 

Fix, J. J. 2004. Technical Basis Document for the Hanford Site – Occupational External Dose, 

ORAUT-TKBS-0006-6, Revision 01, January 9, 2004, Oak Ridge Associated Universities, Oak 

Ridge, Tennessee. 

Francke, H.C., Ludwig, R.L., Roberts, Jr., E., West, C.M., 1966. Memorandum from H.C. 

Francke et. al. to L.E. Burkhart et. al., Np237-U238

 Alloy, Union Carbide Nuclear Division, 

February 19, 1966 1996 pg. 69 BA Neptunium

Frieboth, C.J, Gibbs, F., 2000. Disposition of U-233 in Plutonium Metal and Oxide at the Rocky 

Flats Environmental Technology Site, IT Corporation, RFP-5332, Rev. 1, March 1, 2000. 

Frome, E. L. and P. G. Groer, 2004. Technical Information Bulletin: Individual Dose 

Adjustment Procedure for Y-12 Dose Reconstruction, ORAUT-OTIB-0013, Revision 00, 

September 9, 2004, Oak Ridge Associated Universities, Oak Ridge Tennessee. 

Gilbert, R. 1987. Statistical Methods for Environmental Pollution Monitoring. New York: Van 

Nostrand Reinhold. 

Groer, P. G., 2004. Technical Information Bulletin: Bayesian Methods for Estimation of 

Unmonitored Y-12 External Penetrating Doses with a Time-Dependent Lognormal Model, 

ORAUT-OTIB-0015, Revision 00, September 9, 2004, Oak Ridge Associated Universities, Oak 

Ridge Tennessee. 

Gupton, E.D. 1978. Dosimetry with the ORNL Badge, ORNL/TM-6357, Oak Ridge National 

Laboratory, Oak Ridge, TN. 

Herken, G. 2002. Brotherhood of the Bomb, Henry Holt and Company, New York, (2002), p 

104) 

International Commission on Radiological Protection (ICRP), 1982a. Protection of the Patient 

in Diagnostic Radiology, ICRP Publication 34, Annals of the ICRP, Vol. 9(2/3), 1-82, 1982. 

International Commission on Radiological Protection (ICRP), 1990. Recommendations of the 

International Commission on Radiological Protection, ICRP Publication 60, Annals of the ICRP, 

Vol. 21, Pergamon Press, Oxford, England. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

113 of 203 

International Commission on Radiological Protection (ICRP), 1994. Human Respiratory Tract 

Model for Radiological Protection, ICRP Publication 66, Pergamon Press, Oxford, England 

1994. 

International Commission on Radiological Protection (ICRP). 1994. Dose Coefficients for 

Intakes of Radionuclides by Workers, ICRP Publication 68. Annals of the ICRP, 24(4). 

Tarrytown, NY: Elsevier Science, Inc. 

International Commission on Radiological Protection (ICRP), 1996. Conversion Coefficients for

Use in Radiological Protection Against External Radiation, ICRP Publication 74, Pergamon 

Press, Oxford, England, 1996. 

Ijaz, T. and Adler, T.C. 2004. Technical Basis Document for the Y-12 National Security 

Complex – Occupational Environmental Dose ORAUT-TKBS-0014-4, Rev. 00 PC-2, Oak 

Ridge Associated Universities, Oak Ridge, Tennessee, September 9, 2004. 

International Commission on Radiological Protection (ICRP) 1996. Conversion Coefficients for 

use in Radiological Protection against External Radiation, ICRP Publication 74. Annals of the 

ICRP, 26 (3/4). Tarrytown, NY: Elsevier Science, Inc. 

Jessen, K. 2005. Technical Basis Document for the Y-12 National Security Complex – Site 

Description ORAUT-TKBS-0014-2, Rev. 01, Oak Ridge Associated Universities, Oak Ridge, 

Tennessee, January 19, 2005. 

Kathren, R.L., Shockley, V.E., Thomas, E.M., and Taulbee, T.D. 2003. Dose Reconstruction 

from Occupationally Related Diagnostic X-ray Procedures, ORAUT-TIB-0006, Oak Ridge 

Associated Universities, Oak Ridge, TN, 2003. 

Kerr, G. D. 2003. Technical Basis Document for the Y-12 National Security Complex –

Occupational External Dosimetry, ORAUT-TKBS-0014-6, Rev. 00, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee, November 19, 2003. 

Kerr, G.D. and Smith, M.H. 2004. Accounting for Incomplete Personnel Monitoring Data on 

Penetrating Gamma-Ray Doses to Workers in Radiation Areas at the Oak Ridge Y-12 Plant 

Prior to 1961, ORAUT-PROC-0042, Rev. 00, Oak Ridge Associated Universities, Oak Ridge, 

TN., September 9, 2004. 

Kerr, G. 2005a. Historical Evaluation of the Film Badge Dosimetry Program at the Y-12 

Facility in Oak Ridge Tennessee: Part 1 – Gamma Radiation. ORAUT-RPRT-0032, Revision 

00, May 13, 2005, Oak Ridge Associated Universities, Oak Ridge, Tennessee. 

Kerr, G. 2005b. Historical Evaluation of the Film Badge Dosimetry Program at the Y-12 

Facility in Oak Ridge Tennessee: Part 2 – Neutron Radiation, ORAUT-RPRT-0033, May 13, 

2005, Oak Ridge Associated Universities, Oak Ridge, Tennessee. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

114 of 203 

Klein, J.A. and Turner, D.W., 1994. Issues Relating to Spent Nuclear Fuel Storage on the Oak 

Ridge Reservation, Oak Ridge National Laboratory, CONF-941207-20, December 13–16, 1994. 

Leonwich, J.A., Bihl, D.Fix, J.,Shiplier, D, Napier, B.Rolfes, M. Taulbee, T.,2005. Technical 

Basis Document for the Atomic Energy Operations at the Iowa Army Munitions Plant, ORAUTTKBS-0018, Rev. 1, March 14, 2005. 

Livingston, R.S. and Zucker, A., 1961. Electronuclear Division Annual Progress Report for 

Period Ending December 31, 1961, TID-4500, Oak Ridge National Laboratory, Oak Ridge, 

Tennessee. 

Leverett, W. C. Letter to E. P. Wigner, Hazards of Piles Y-12, January 9, 1947. 

Ludlum Measurements, Inc. 2000. Response Curves

<http://www.ludlums.com/RespCurvHtm/RespCurv-6.htm> (June 15, 2005) 

McLendon, J.D., 1960. Letter from J.D. McLendon to G.W. Flack dated August 25, 1960. 

“Personnel Radiation Exposure Data,” Union Carbide Nuclear Company, Oak Ridge, Tennessee. 

McLendon, J.D. Letter of 21 May 1962 to L.E. Burkhart (Page 27 of 32 in Database 010002413, 

scanned 12-15-03). 

McLendon, J. D., 1963. Memorandum from J.D.McLendon to distribution, Y-12 Plant Quarterly 

Health Physics Report, Union Carbide Nuclear Division, Oak Ridge, Tennessee, February 7, 

1963. 

McLendon, J. D., 1972. Memorandum from J.D. McLendon to H.E. Alvey et. al. Uranium 

Urinalysis Program, Union Carbide Nuclear Division, Oak Ridge, Tennessee, January 10, 1972. 

Merwin, S.E. 2003. Dose Reconstruction Project for NIOSH – External Dose Reconstruction, 

ORAUT-PROC-0006, Rev. 00, PC-2, December 11, 2003. 

Murray, W. E. 2003. Technical Basis Document for the Y-12 National Security Complex – 

Occupational Medical Dose, ORAUT-TKBS-0014-3, Rev. 00, PC-1, September 9, 2004. 

Murray W. E. 2004a. Technical Basis Document for the Y-12 National Security Complex – Y-12 

Site Profile ORAUT-TKBS-0014-1 Rev. 00, PC-1, Oak Ridge Associated Universities, Oak 

Ridge, Tennessee, November 24, 2004. 

Murray, W.E. 2004b. Technical Basis Document for the Y-12 National Security Complex – 

Occupational Medical Dose, ORAUT-TKBS-0014-3, Rev, 00, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee, September 9, 2004. 

National Council on Radiation Protection and Measurement (NCRP). 1976. Environmental 

Radiation Measurements, NCRP Report No. 50. Bethesda, MD. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

115 of 203 

National Council on Radiation Protection and Measurements (NCRP), 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. 

National Institute for Occupational Safety and Health (NIOSH) 2002. External Dose 

Reconstruction Implementation Guide, OCAS-IG-001, Rev. 1, August 2002. 

Neton, J.W. 2003. Neutron Exposures at the Savannah River Site, OCAS-OTIB-007, Office of 

Compensation Analysis and Support, Cincinnati, Ohio, September 17, 2003.

Neton, J.W. 2004. Estimation of Ingestion Intakes, OCAS-TIB-009, Rev. 00, Office of 

Compensation Analysis and Support, April 13, 2004. 

Norris, R.S. 2004. Dismantling U.S. Nuclear Warheads, Bulletin of Atomic Scientists, 

January/February 2004, pp. 72-74. 

Oliver, R.W., 1997. Transuranic Hazards Assessment at the Y-12 Plant, RCO/TR-001, 

Lockheed Martin Energy Systems, Inc., Oak Ridge, Tennessee, July 1997. 

Oliver, R.W. 2001. Atypical Radionuclide Assessment at the Y-12 National Security Complex, 

BWXT Y-12, Oak Ridge, Tennessee, May 17, 2001. 

Owen, F.E., 1964. Beta-Gamma Dose Rates From U232 in U233

, HW-81964, Hanford Atomic 

Products Operation, Richland, Washington. 

Patterson, G.R., Jr., West, C.M., McLendon, J.D., 1957. The Y-12 Health Physics Program, Y1186, Union Carbide Nuclear Company, Oak Ridge, Tennessee, November 1, 1957. 

Patterson, H.W., Thomas, R.H., 1973. Accelerator Health Physics, Academic Press. 

Polednak A.P, Keane, A.T., and Beck, W., 1982. Estimation of radiation doses to the lungs of 

early uranium processing plant workers, Environ Res. August 1982 (2):313-28. 

Rich, B. L. and Chew, M.H. 2005. Technical Basis Document for the Y-12 National Security 

Complex –Occupational Internal Dose, ORAUT-TKBS-0014-5, Rev. 01, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee, May 10, 2005. 

Rollins, E.M. 2004. Technical Information Bulletin – Maximum Internal Dose Estimates for 

Certain DOE Complex Claims, ORAUT-OTIB-0002, Rev. 01, January 10, 2004, Oak 

RidgeAssociated Universities, Oak Ridge Tennessee. 

Rollins, E. M. 2004. Technical Basis Document for the Nevada Test Site – Occupational 

External Dose, ORAUT-TKBS-0008-6, Rev. 00, September 21, 2004, Oak Ridge Associated 

Universities, Oak Ridge Tennessee. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

116 of 203 

Sanders, M., 1963. Letter M. Sanders to G.W. Evans et. al., “Thorium Contamination Control,” 

Union Carbide Nuclear Company, Y-12 Plant, Oak Ridge, TN., March 12,1963. (OUO – Privacy 

Act) 

S. Cohen & Associates (SC&A) 2004. SC&A Standard Operating Procedure for Performing 

Site Profile Reviews, Rev. 0, March 18, 2004. 

Savignac, N. 2003. Technical Basis Document for the Hanford Site – Occupational 

Environmental Dose, ORAUT-TKBS-0006-4, Revision 00, October 15, 2003, Oak Ridge 

Associated Universities, Oak Ridge, Tennessee. 

Scalsky E. D 2003. Technical Basis Document for the Hanford Site – Occupational Medical 

Dose, ORAUT-TKBS-0006-3, Revision 00, September 16, 2003, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee. 

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. 

Scott, L.M., 1979. Memorandum: Remarks on Personnel Dosimetry Monitoring at Y-12. 

Snapp, L.M., 1995. Technical Basis Document for the Internal Dosimetry Program at the Y-12 

Plant, Y/DQ-40, Revision 1, Lockheed Martin Energy Systems, Inc., October 27, 1995, 

Chapter 3. 

Steckel, L.M. and West, C.M., 1966. Characterization of Y-12 Uranium Process Materials 

Correlated with In Vivo Experience. Union Carbide Corporation, Oak Ridge, Tennessee, 

July 28, 1966. 

Struxness E. G. 1949. Health Physics Report; January 1-31, 1949, Union Carbide Corporation, 

Oak Ridge, Tennessee, February 14, 1949. 

Struxness, E. G. 1951. Health Physics Progress Report; July 1-December 31, 1951, Union 

Carbide Corporation, Oak Ridge, Tennessee, December 31, 1951. 

Struxness, E. G. 1952. Health Physics Progress Report; July 1-December 31, 1951, Union 

Carbide Corporation, Oak Ridge, Tennessee, March 15, 1952 

Struxness, E. G. 1953. Health Physics Progress Report; January 1, 1952-July 1, 1952, Union 

Carbide Corporation, Oak Ridge, Tennessee, February 19, 1953. 

Thomas, J.M. 1995. Email: RadCon Requirements for Neptunium Processing, June 12, 1995. 

Thomas, J.M. and Bogard, R.S. 1994. Technical Basis for Beta Skin Dose Calculations at the 

Y-12 Plant, Martin Marietta Energy Systems, Inc., Oak Ridge, Tennessee, March 1994. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

117 of 203 

Union Carbide Nuclear Company (UCNC) 1959. In Vivo Conversion Factor (U), Memorandum 

dated June 2, 1959. 

U.S. Atomic Energy Commission (AEC) 1948. Advisory Committee For Biology and Medicine, 

Thirteenth Meeting, December 10-11, 1948, U.S. Department of Energy Archives, 1947-51 

Secretariat Files, Box 1217, Folder 337, Germantown, MD. 

U.S. Atomic Energy Commission (AEC) 1949. Advisory Committee for Biology and Medicine, 

16th Meeting, June 11, 1949, Transcript, U.S. Department of Energy Archives, AEC Division of 

Biology and Medicine Collection, Box 3218, Folder ACBM Meeting, Germantown, MD. 

U. S. Atomic Energy Commission (AEC) 1965. U. S. Atomic Energy Commission Investigation 

of X-ray Exposure Incident at Y-12, U.S. AEC Report of Investigating Committee, April, 28, 

1965. 

U.S. Department of Energy (DOE) 1984. Health Physics Appraisal of Y-12 Plant, Oak Ridge, 

Tennessee, FY 1984, Oak Ridge Operations Office, Oak Ridge, Tennessee. 

U.S. Department of Energy (DOE) 1985. The Report of the Joint Task Force on Uranium 

Recycle Materials Processing, Oak Ridge Operations, Report No. DOE/OR-859, September 

1985. 

U.S. Department of Energy (DOE) 1988. Technical Safety Appraisal of Y-12 Plant 4/25/88 thru 

5/6/88, Volume 1, Oak Ridge Operations Office, Oak Ridge, Tennessee. 

U.S. Department of Energy (DOE) 1994a. Plutonium Working Group Report on Environmental 

Safety and Health Vulnerabilities Associated with the Storage of Plutonium, Volume 2, Part 9, 

Oak Ridge Site Working Group Assessment Team Report, DOE/EH-0415, September 1994. 

U.S. Department of Energy (DOE) 1994b. Plutonium Working Group Report on Environmental, 

Safety and Health Vulnerabilities Associated with the Storage of Plutonium, Volume 2, Part 9, 

Oak Ridge Site Working Group Assessment Team Report, DOE/EH-0415, September 1994. 

U.S. Department of Energy, Defense Nuclear Facility Safety Board, 1995. Status of Highly 

Enriched Uranium Processing Capabilities at Building 9212 Oak Ridge Y-12 Plant, Technical 

Report No. 9, December 8, 1995. 

U.S. Department of Energy (DOE) 1996. Highly Enriched Uranium Working Group Report on 

Environmental, Safety and Health Vulnerabilities Associated with the Department’s Storage of 

Highly-Enriched Uranium, Volume II, Number 1, Oak Ridge Y-12 Working Group and Site 

Assessment Teams Reports, DOE-EH-0525, December 1996. 

U.S. Department of Energy, Defense Nuclear Facility Safety Board, 1997. Technical Report No. 

13, Uranium-233 storage Safety at Department of Energy Facilities, February 1997. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

118 of 203 

U. S. Department of Energy (DOE) 1999. DOE Order 5400.11 and later DOE-STD-1098/99 

Radiological Control, 1999.

U.S. Department of Energy (DOE) 1998a. Defense Nuclear Facility Safety Board, Oak Ridge 

Weekly report, October, 2, 1998. 

U.S. Department of Energy (DOE) 1998b. Defense Nuclear Facility Safety Board, Oak Ridge 

Weekly report, October 30, 1998. 

U.S. Department of Energy (DOE) 2000a. Defense Nuclear Facility Safety Board, Oak Ridge 

Weekly report, March 10, 2000. 

U.S. Department of Energy (DOE) 2000b. Recycled Uranium Mass Balance Project Y-12 

National Security Complex Site Report, Y/LB-16,036, December 2000. 

U.S. Department of Energy, (DOE) 2000c. Office of Environment, Safety and Health, Exposure 

Assessment Project at the Paducah Gaseous Diffusion Plant, December 2000. 

U.S. Department of Energy (DOE) 2000. Recycled Uranium Mass Balance Project Y-12 

National Security Complex, Y/LB- 16,036, Rev. 1, December 2000. 

U.S. Department of Energy (DOE) 2002a. Office of Inspector General, Depleted Uranium 

Operations at the Y-12 National Security Complex, Audit Report, DOE/OIG-0570, September 

2002. 

U.S. Department of Energy (DOE) 2002b. Defense Nuclear Facilities Safety Board, oak Ridge 

Site Representative weekly report, March 22, 2005) 

U.S. Department of Energy (DOE) 2005a. Defense Nuclear Facilities Safety Board, Oak Ridge 

Site Representative weekly report, March 22, 2005. 

U.S. Department of Energy (DOE) 2005b. Defense Nuclear Facilities Safety Board, Oak Ridge 

Site Representative weekly report, April 15, 2005. 

Vallario, E.J., Hankins, D.E., and Unruh, C.M. 1969. AEC Workshop on Personnel Neutron 

Dosimetry, BNWL-1616, Pacific Northwest Laboratory, Richland, Washington. 

Watkins, J. P., Cragle, D. L, Frome, E. L., West , C. M, Crawford-Brown, D. J.and Tankersley, 

W. G. 1994. Adjusting External Doses from the ORNL and Y-12 Facilities for the Oak Ridge 

Nuclear Facilities Mortality Study, Oak Ridge Institute for Science and Education, Center for 

Epdeimiologica Research,, http://www.orau.gov/cer/reports.htm#Adjusting. 

Watkins, J. P., et al. 2004. Historical Evaluation of the Film Badge Dosimetry Program at the 

Y-12 Facility in Oak Ridge Tennessee, Part 1, Gamma Radiation, ORAU Technical Report 

2004-0888, December 30, 2004. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

119 of 203 

Watson, J. E. Jr., J. L. Wood, W. G. Tankersley, and C. M. West 1994. Estimation of Radiation 

Dose for Workers Without Monitoring Data for Retrospective Epidemiologic Studies, Health 

Physics (The Radiation Safety Journal), volume 67, number 4, pp. 402-405. 

West, C.M., 1954a, Letter from C.M. West to H.W. Saylor, “July Health Physics Report – 

Chemical and Foundry Areas Bldg. 9212 (U)”, Carbide and Carbon Chemical Company, Y-12 

Plant, Oak Ridge, TN. 

West, C.M., 1954b, Letter from C.M. West to H.W. Saylor, “August Health Physics Report – 

Chemical and Foundry Areas Bldg. 9212 (U)”, Carbide and Carbon Chemical Company, 

September 13, 1954. 

West, C.M., 1954c, Letter from C.M. West to H.R. Gwinn, “Health Physics Report – Building 

9204-3, “Carbide and Carbon Chemical Company, Y-12 Plant, Oak Ridge, TN., September 7, 

1954. (OUO – Privacy Act) 

West, C.M., 1956a, Memorandum to F. Williams, Building 3995, “Change in the Recording of 

Minimum Detectable Badges,” Carbide and Carbon Chemical Company, June 21, 1956. 

West, C. M., 1956b, Letter from C.M. West to W.A. Pfeiler et. al., “Recalculation of Urine 

Results,” Union Carbide Company, Y-12 Plant, Oak Ridge, TN, August 30, 1956. (OUO – 

Privacy Act) 

West, C.M., 1961, Memorandum to distribution, “Plutonium Contamination of Beryllium,” 

Carbide and Carbon Chemical Company, September 1, 1961. 

West, C.M. 1965, Health Physics Considerations Associated with Thorium Processing. Y-KB53, Union Carbide Corporation, Oak Ridge, Tennessee, January 19, 1965. 

West, C.M ., 1981, Memorandum to H.L. Fore, “Discrepancy in 1954 Y-12 Film Results,” 

Union Carbide, Nuclear Division, August 18, 1981. 

West, C.M., 1985, Memorandum to M.L. Jones, Building 9106, “New Uranium Air Limits and 

Their Impact on Y-12,” Union Carbide, Nuclear Division, January 28, 1985. 

West, C. M., 1991, Correction of Y12 Film Badge Results in the 1950 Through 1955 Period, Oak 

Ridge Associated Universities, Oak Ridge, Tennessee, February 19, 1991. 

West, C.M., 1993, Brief History of the Y-12 External Monitoring Program, Oak Ridge Institute 

for Science and Education (ORISE), June 7, 1993. 

West, C.M. and Rogers, Jr., E., 1962. Health Physics Experience During Processing of U-233 in 

Y-12 Plant, Y-KB-19, September 12, 1962. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

120 of 203 

Wilson, R.H., Fix, J.J., Baumgartner, W.V., and Nichols, L.L., 1990, Description and Evaluation 

of the Hanford Personnel Dosimeter Program From 1944 Through 1989, PNL-7447, Pacific 

Northwest Laboratory, Richland, Washington. 

Wollan, R.O., 1954, Letter from R.O. Wollan to B. Harmatz, “Uranium and Plutonium 

Urinalysis Report for Bldg. 9204-3,” Carbide and Carbon Chemical Company, Y-12 Plant, Oak 

Ridge, TN., June 18, 1954. (OUO – Privacy Act) 

Y-12 Plant, 1963, Y-12 Plant Quarterly Health Physics Report – Fourth Quarter CY 1962, 

Y/KB-26, Oak Ridge, Tennessee. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

121 of 203 

ATTACHMENT 1: REVIEW OF THE NIOSH SITE PROFILE FOR 

THE Y-12 NATIONAL SECURITY COMPLEX 

Presentation to the Advisory Board on Radiation and Worker Health 

Joseph Fitzgerald 

July 5, 2005 

St. Louis, Missouri 

Y-12 Review: Background

• Advisory Board expedited Y-12 review at April 27 meeting to support review of SEC petitions

• SC&A review conducted May 3-June 30, 2005, including classified records and onsite interviews

• Focused on current ORAU team TBDs

• Three site visits to Y-12

• All interview notes, doc reviews, and draft report reviewed by

derivative classifier at Y-12

• Conference calls held between TBD reviewers and ORAU authors

• Report to be submitted late-July

Y-12 Review: TBDs Reviewed

• Y-12 Site Profile (Introduction)

• Site Description

• Occupational Medical Dose

• Occupational Environmental Dose

• Occupational Internal Dose

• Occupational External Dose
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

122 of 203 

Y-12 Review: Supporting Documents Reviewed

• Technical Information Bulletin – Occupational Dose from Elevated Ambient 

Levels of External Radiation 

• Technical Information Bulletin: Individual Dose Adjustment Procedure for Y12 Dose Reconstruction 

• Technical Information Bulletin: Bayesian Methods for Estimation of 

Unmonitored Y-12 External Penetrating Doses with a Time-Dependent

Lognormal Model

• Internal Dosimetry Coworker Data for Y-12

• Historical Evaluation of the Film Badge Dosimetry Program at the Y-12 

Facility in Oak Ridge Tennessee: Part 1 – Gamma Radiation

• Historical Evaluation of the Film Badge Dosimetry Program at the Y-12 

Facility in Oak Ridge Tennessee: Part 2 – Neutron Radiation

• Accounting for Incomplete Personal Monitoring Data on Penetrating 

Gamma-Ray Doses to Workers in Radiation Areas at the Oak Ridge Y-12 

Plant Prior to 1961 

Y-12 Review: Site Description

• Operating contractors: TEC, 1943-1947; Union Carbide, 1947-1984; 

MMES/LMES, 1984-1998; BWXT, 1998-present.

• Construction contractors: Rust Engineering, MK Ferguson, Bechtel 

Jacobs

• Secure site with 531 buildings on 811 acres

• Worker population variable over time: 4,700 employees today;

11,000 during peak production; about 40,000 during TEC years
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

123 of 203 

Y-12 Missions

• Electromagnetic separation of uranium (1943-1947)

• Production of secondaries and cases

• Refurbishment and dismantling of nuclear weapons components

• Storage and processing of uranium and lithium materials and parts

• Production support to weapons laboratories

• DOD and NASA program support

Y-12 Review: TBD Strengths

• Internal dosimetry TBD undergoing revision to include a more

extensive treatment of radionuclides other than uranium

• Y-12 internal and external monitoring techniques detailed

• Beta exposure under investigation as a complex wide issue

• “Radiation generating devices” TIB under development.

• Recent records retrieval efforts have been more comprehensive

• TBD is supplemented with a number of TIBs and reports that 

provide further direction to the dose reconstructor

• Improved approach in addressing recycled uranium, compared with 

other TBDs reviewed
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

124 of 203 

Y-12 Historic Radiation Protection Program for 

Uranium 

Y-12’s HP program had sound features and conducted extensive monitoring, but fell short for uranium:

• Eating, drinking, smoking permitted in uranium production areas until 1988; no egress monitoring (DOE, 1988; DNFSB, 1993)

• Lack of formal worker radiation protection training (DNFSB, 1993)

• Inadequate in-plant radioactive contamination zoning (DOE, 1988; DNFSB, 1993)

• Limited shielding for beta/gamma radiation from depleted uranium (DOE, 1988)

• Respirators were available but discretionary for all but a few high exposure jobs; supervisors often decided (DOE, 1988)

• Plant action levels exceeded frequently (sometimes orders of magnitude) without

lasting remedies (Y-12 Quarterly HP reports, 1960s era)

• Line programs unresponsive to identified radiation protection deficiencies (DOE, 1988)

• Large backlog of uranium materials, some remaining in work areas for several 

decades

Y-12 site expert/workers interviews

• Interviews with:

– 8 BWXT environment, safety and health personnel

– 25 BWXT production and maintenance personnel

– 3 Wackenhut Services, Inc (WSI) security guards

• Employment histories back to 1969

• Key comments (corroborated by multiple workers):

– Workers moved often between operations

– Workers employed at more than one Oak Ridge site during career

– Overtime common

– Y-12 support workers (e.g., maintenance, janitors) serviced ORNL operations at Y-12 and 

entire plant area

– Plutonium, 233U, and thorium present at the facility as a process material and contaminant

– Plutonium contamination exists in sealed gloveboxes

– Fecal analysis replaced urinalysis belatedly in 1998 for insoluble uranium oxides

– Selective breathing zone sampling was utilized in the work area

– Management compelled workers to remain at work stations; number of instances of 

production taking priority over safety in past history
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

125 of 203 

Y-12 Review: Available Radiation Monitoring Data

• High uranium air concentrations and no external monitoring data for

1943-1947

• Less than 25% of workers monitored externally from 1948-1960

• No external monitoring data readily available for 1948-1950

• No urinalysis data located prior to 1948

• Lung counting initiated in 1958 as a result of thorium production

• Uranium urinalysis initially limited to production workers, but 

expanded to include other workers starting in 1954

• Significant reduction in urinalyses (whole departments) in 1972 with

increased reliance on lung counting

• Limited tritium and plutonium bioassay initiated in 1957

• Routine fecal sampling program initiated in 1998

Y-12 findings

• Y-12 TBDs need to clearly address support workers who 

were not routinely badged before 1961, nor bioassayed

before 1994

– Maintenance and janitorial employees, and some inspectors and 

material inventory handlers not monitored routinely, yet had full

access to Y-12 radiological control areas

– Exposed to HEU, insoluble uranium oxides, thorium, TRU, 

ORNL neutron sources

– Work included cleaning contaminated areas, fixing contaminated 

process equipment, and moving material.

– No records apparently exist regarding work location assignments
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

126 of 203 

Y-12 findings

• TBDs need to better characterize “incidental” sources of

chronic workplace exposure

– Uranium chip fires were relatively frequent incidents (perhaps 

several times a shift) that filled operating areas with uranium

fumes

– Exhaust fans sometimes turned off in operating areas leading to 

back-flow of contaminated air through ducts

– Exposure from radon and radium preferentially vaporized from

molten uranium and thorium metal in casting operations

– Absence of formal incident documentation

Y-12 findings

• Internal dose TBD incomplete in its treatment of historic

dose contributions of radionuclides other than uranium

– Y-12 handled 3H, 90Sr, 99Tc, 210Po, 228Th, 232Th, 239Pu, 241Pu, 

237Np, 233U and 241Am

– Some used in experimental activities, others in production as

source material or recycled feed contaminant 

– Thorium and neptunium processing particularly important

– Plutonium involved with “sputtering” of components during 

production; also as contaminant in LANL/RF “returns”

– Y-12 support workers likely exposed to nuclides associated with 

ORNL operations

– NIOSH recognizes issue and ORAU has ongoing assessment of

thorium, 233U, and recycled uranium
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

127 of 203 

Y-12 findings

• Potential for missed dose at Y-12 should be more

completely addressed

– Unmonitored categories of workers

– External doses to recycled uranium workers prior to 1961

– Uncertainties and detection limits of bioassay techniques

– Radioactive material solubility and particle size assumptions

– Ingestion pathway

– Application of co-worker data

– Adequacy of bioassay techniques

• Frequency of sampling vs. type of material

• Appropriateness of bioassay techniques for insoluble uranium 

oxides

Y-12 findings

• External Dose - Neutron radiation fields not defined sufficiently:

– Potential neutron fields not adequately addressed to validate 

NTA film response; mostly hard-neutron spectra data by PNL 

were used. Other sources that need to be considered are:

• Spontaneous fission neutrons

• Moderated (α, n) sources in solutions/compounds

• Subcritical/critical assemblies

• Moderated neutrons from the 86” cyclotron
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

128 of 203 

Y-12 findings

• External Dose - Radiation fields from radiation generating devices 

(RGDs) need to be addressed:

– By 1950, 10 non-medical x-ray machines were installed at Y-12, ranging from 40 kVp, 16 mA, to 1 MeV, 10 mA

– Not apparent that appropriate workers were monitored with the 

correct dosimeters as RGDs became more common in late 

1940s and early 1950s

– 86” cyclotron received little mention, but was one of largest 

contributors to neutron dose at Y-12 (ORAU-RPRT-0033)

– Other potential radiation exposure sources include 10 Ci 60Co

unit and experimental pile assembly

– Exposures from these sources were real possibility (i.e., overexposure to radiographic x-ray unit in 1965)

Y-12 findings

Prior-to-1961 dose assignments:

• The methods outlined in the TBD are valid if the pre-1961 badged

workers were the highest exposed workers in the work force

• Selective bias due to administrative criteria that assigned badges and 

type of exposures, a priori, is confounding factor

• The methods outlined in the TBD are not claimant favorable if the pre1961 badging was effectively cohort badging of various groups; this 

would represent average exposure only if the badged individuals were

truly representative of the cohort

• No documentation has been found for methods used at Y-12 to ensure

that the most highly exposed workers were monitored, such as by

random monitoring or temporary 100% monitoring
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

129 of 203 

Y-12 findings

Potential missed internal dose associated with intake model:

• Intake model needs to consider Type F uranium compounds in

addition to Types M and S compounds

• Intake model needs to consider different particle sizes present at 

Y-12

• Intake model needs to consider ingestion pathways, in addition to

inhalation

• Intake model needs to address uncertainties in bioassay

measurements and capacity to detect intakes from urine samples 

after exposure to Type S uranium compounds

• Intake model needs to reflect 48 hour delay in obtaining urine 

sample from worker

• Intake model needs to consider more claimant favorable sampling

distribution (e.g., 95th percentile) from co-workers database

Y-12 Findings

• Occupational Environmental Dose TBD should consider additional

pathways

– Inadvertent ingestion of radioactive materials

– Inhalation of radionuclides other than uranium (e.g., thorium)

– Airborne uranium and other radionuclides from off-site sources

(e.g., ORNL)

– Exposure to uranium fumes from burning of DU chips
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

130 of 203 

Y-12 findings

• Extremity and skin doses should be given attention in the 

TBDs

– Uranium metal handling has associated high exposure of hands

and arms

– 232Th similarly has beta-emitting progeny that present 

radiological hazard for direct handling 

– Average quarterly skin dose (Bldg 9206, 1961-1962):

• Mechanical operators: 100-180 mrem N = 1750

• Product certification: 220-320 mrem N = 451

• Chemical operators: 230-370 mrem N = 523

– Use of ring dosimeters limited to specialized activities

– ORAU team review of issue ongoing

Conclusions

• Review found that the Y-12 site profile to be an adequate accounting

of the “core” uranium exposure and dosimetric history of the plant. 

However, the TBDs can be improved (and work is already underway

by the ORAU team) to better characterize the significance and 

implications of actual operational and dosimetric practice. 

• Particular attention is needed for:

– Adequacy of bioassay program for detection of insoluble uranium 

oxides, acute uptakes, and radionuclides other than uranium

– Past bioassay practice for high-fired uranium oxides

– Use of co-worker external dose assignments before 1961

– Spectral field basis for assumptions on NTA threshold assumptions

– Ingestion pathway

– Unmonitored or intermittently monitored workers routinely working in

production areas

– Adequacy and completeness of environmental dose methodologies
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

131 of 203 

Acknowledgment

The Y-12 SC&A review team wishes to acknowledge 

and thank the following individuals for support and

cooperation during this expedited review:

Stu Hinnefeld, NIOSH

Bill Murray, ORAU

Janet Wood & Steve Wylie, BWXT/Y-12
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

132 of 203 

ATTACHMENT 2: NIOSH TECHNICAL DOCUMENTS

CONSIDERED DURING THE REVIEW 

Technical Basis Documents 

ORAUT-TKBS-0014-1, Technical Basis Document for the Y-12 National Security Complex – 

Y-12 Site Profile (Murray 2004a) 

ORAUT-TKBS-0014-2, Technical Basis Document for the Y-12 National Security Complex – 

Site Description (Jessen 2005) 

ORAUT-TKBS-0014-3, Technical Basis Document for the Y-12 National Security Complex – 

Occupational Medical Dose (Murray 2004b) 

ORAUT-TKBS-0014-4, Technical Basis Document for the Y-12 National Security Complex – 

Occupational Environmental Dose (Ijaz and Adler 2004) 

ORAUT-TKBS-0014-5, Technical Basis Document for the Y-12 Site –Occupational Internal 

Dose (Rich and Chew 2005)

ORAUT-TKBS-0014-6, Technical Basis Document for the Y-12 Site –Occupational External 

Dosimetry (Kerr 2003) 

Technical Support Documents 

ORAUT-OTIB-0013, Technical Information Bulletin: Individual Dose Adjustment Procedure 

for Y-12 Dose Reconstruction (Frome and Groer 2004) 

ORAUT-OTIB-0015, Technical Information Bulletin: Bayesian Methods for Estimation of 

Unmonitored Y-12 External Penetrating Doses with a Time-Dependent Lognormal Model, 

(Groer 2004) 

ORAUT-OTIB-0029, Internal Dosimetry Coworker Data for Y-12, (Brackett 2005) 

ORAUT-RPRT-0032, Historical Evaluation of the Film Badge Dosimetry Program at the Y-12 

Facility in Oak Ridge Tennessee: Part 1 – Gamma Radiation (Kerr 2005) 

ORAUT-RPRT-0033, Historical Evaluation of the Film Badge Dosimetry Program at the Y-12 

Facility in Oak Ridge Tennessee: Part 2 – Neutron Radiation (Kerr 2005) 

ORAUT-PROC-0042, Accounting for Incomplete Personal Monitoring Data on Penetrating 

Gamma-Ray Doses to Workers in Radiation Areas at the Oak Ridge Y-12 Plant Prior to 1961

(Kerr and Smith 2004)
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

133 of 203 

ATTACHMENT 3: KEY QUESTIONS FOR NIOSH/ORAU 

REGARDING SITE PROFILE DOCUMENTS

SITE DESCRIPTION (ORAUT-TKBS-0014-2) 

1. During the course of your preparation of the TBD, how extensive was your review of 

classified documents pertinent to dose reconstruction? Which documents were reviewed?

2. The TBD states that Y-12 manufactured nuclear weapons components. Can you 

elaborate, on an unclassified basis, what general types of components were manufactured, 

handled, or processed? 

3. Page 5: It is likely that classified information will be relevant to reconstructing doses at 

an individual claimant level. How will this information be redacted in a way that allows 

the dose reconstructions to be adequately scrutinized? 

4. Does NIOSH have any estimates of and historical timelines for nuclear materials flow

(i.e., uranium-233, recycled uranium, and other nuclear materials) from production and 

AWE sites going to Y-12 for storage, processing, and shipment? 

5. Given the historical relationship between the Y-12 Laboratory and Oak Ridge National 

Laboratory, particularly between 1943 and the early 1970s, has NIOSH determined if 

radiation exposure data are accurate and complete for people who cycled in and out of 

these two labs during that time period?

6. A significant fraction of the total estimated amount of recycled uranium generated in the 

DOE complex was processed at the Y-12 plant. Has NIOSH ascertained: (a) what 

processes were likely to concentrate trace radiological contaminants that could be inhaled 

or ingested; (b) if individual personnel dosimetry for Y-12 RU workers is accurate or 

complete; and (c) whether workplace measurements in areas processing recycled uranium

are sufficient to support current assumptions in the TBD?

7. Table 2.3-1, ORAUT-TKBS-0014-2, lists radionuclides of concern by building number. 

What is the meaning of the N/A listed for uranium and 239Pu?

8. Table B.4-3, ORAUT-TKBS-0014-2, lists radiological work permit bioassay indicators 

for routine uranium handling jobs. Is the use of the bioassay indicator specific to 

particular years? If so, which years?

9. There were several operations performed by ORNL personnel at the Y-12 plant. Are the 

individuals associated with these operations considered Y-12 or ORNL workers? Have 

you investigated how these operations may have affected the dose to Y-12 workers?

10. Has NIOSH investigated whether thorium-232 fuel for the Molten Salt reactor or any 

other reactor was ever fabricated at Y-12?
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

134 of 203 

11. Is there a databank for incidents at Y-12?

12. Pages 6 and 7: What other uranium compounds were used and in what contexts? In 

particular, was the use of these other forms widespread?

13. Page 7: Electromagnetic operations beyond 1947 were directed towards research on new 

radionuclides for medical or other uses, such as separation of lithium. This must have 

meant that off-batch processing operations were conducted. Inadequate radiological 

control in individual operations could have resulted in high exposures of the limited 

numbers of workers involved. Are there ways of associating individual workers with 

those operations, determining the radiological controls in force and their implementation, 

and recovering monitoring data relevant to individual dose reconstruction? Also, could 

there have been other off- batch processing operations that could have given rise to 

significant operational exposures, e.g. handling of large amounts of Np-237 recovered at 

plants such as Fernald?

14. Page 7: Although operations were mainly weapons-related, with the plutonium present 

likely being of low “burnup,” it is still surprising that Pu-240 is not listed in Table 2.3-1. 

Here and elsewhere, is Pu-239 actually Pu-239/240?

15. Page 8: Both fusion research and weapons-related activities can involve the handling of 

substantial quantities of tritium. Tritium gets a brief mention in the appendix to this 

document and in other documents, but the operations involving handling of tritium are 

not discussed. Is there any reason for this lack of TBD citation and guidance? Could this 

topic be expanded upon given the available data? 

16. Page 11: From Section 2.4.6, it seems that the new filters may not have operated 

effectively from their installation in 1955 through July 1956. As well as large material 

losses, could high worker exposures have occurred over that period, e.g. in trying to get 

the new filter system to operate efficiently?

17. Page 15: The first assessment of the S-3 ponds indicated that no special precautions were 

required for clean-up work. However, the second survey seems to have indicated that 

precautions were required. Is this correct and was work on capping the ponds undertaken 

without such precautions between the two assessments? If so, have the radiological 

consequences been assessed?

18. Page 19: Two of the default solubility assumptions comprise a mixture of Class S and 

Class M. In interpreting the data, was consideration given to using a single class with a 

solubility intermediate between Classes M and S rather than assuming separate 

contributions from the two classes with different solubilities?

19. Page 20: In Table 2.9-1 not all the uranium compounds are assigned solubility classes. 

Why is this? If ammonium diuranate should not be assigned to a single class (see 

footnote to the table), how should the assignment in the table be interpreted?
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

135 of 203 

20. Page 21: IMBA is used in other technical basis documents reviewed, but for 

Y-12, a DOSE-66 code is used. Why has the DOSE-66 code been given precedence, and 

has a comparison been made between IMBA and DOSE-66 in interpreting the Y-12 

bioassay data?

21. Page 29: In Table B.4-1, why are Pu-238 and Pu-239 assigned to different solubility 

classes? How is this reconciled with Table B.4-2?

22. Page 30: Why is H-3 as vapor first mentioned in Table B.4-2 and not discussed in the 

main text? 

23. Page 30: Table B.4-4 gives average data. Are the proportions of S and M available for 

each individual data point, so that a measure of variability can be obtained? 

OCCUPATIONAL MEDICAL DOSE (ORAUT-TKBS-0014-3) 

1. The TBD (pg. 6) indicates that from 1943 to1947, pre-employment chest x-rays were 

taken with a photofluorographic (PFG) unit as evidenced by the 4 x 10 inch films found 

in medical records. What is known about the probable dose from these chest x-rays? 

What is known about the calibrations of these units?

2. The TBD (pg. 7) indicates that at Y-12, no factors were identified for Type I and II 

equipment and, as such, organ doses were based on assumed technique factors, and that 

these were developed on the basis of x-ray techniques contemporary with the time period 

1943–1968. Since the x-ray technique factors may not be reliable and default values for 

entrance kerma have been developed for use in calculating organ dose conversion factors, 

what can you tell us about the default values for the three most commonly used 

occupational medical diagnostic x-ray procedures, i.e., PA, lateral, and PFG chest films? 

3. Did Y-12 perform lateral chest x-rays? If so, how were these doses calculated?

4. How was sufficient conservatism built into the determination of default values to ensure

with near certainty (>99% confidence) that the actual exposures from the specified 

procedures mentioned in ORAUT-TKBS-0014-3 would not exceed these default values, 

thus ensuring claimant favorability? 

5. What was the skin-to-surface distance assumed for occupational medical x-rays (i.e., 

photofluorography and standard chest x-rays)?

OCCUPATIONAL ENVIRONMENTAL DOSE (ORAUT-TBKS-0014-4) 

1. ORAU addresses only two exposure pathways: inhalation of uranium in ambient air and 

external exposure to direct penetrating radiation. Whenever individuals perform outdoor 

work, there is a potential for ingestion of soil and other windblown particulates that are 

larger than normally respirable particles. This pathway is traditionally included in dose
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

136 of 203 

assessments (see EPA “Exposure Factors Handbook”). What is the basis for not 

including inadvertent ingestion of radioactively contaminated soil and other finely 

dispersed radioactive materials?

2. There are several references in this TBD to monitoring of fission products. What is the 

basis of not including consideration of inhalation of radionuclides other than uranium? 

3. The TBD states that uranium air concentrations were not available after 1999. However, 

since TDEC continued to collect samples and analyze them for gross α and gross β, 

uranium air concentrations could be estimated from these measurements from the ratios 

of uranium concentrations to the gross α and gross β measurements made in earlier years. 

What is the basis for excluding these data from developing the guidance for dose 

reconstruction?

4. Data regarding the AMAD or the chemical form of the airborne uranium is required by 

dose reconstructors, who need to assign dose coefficients to the uranium isotopes. Does 

NIOSH attempt to determine the AMAD and the specific chemical form of uranium?

5. The empirical χ/Q approach is based on several premises which are not explicitly stated 

and which need to be explored and justified before this approach can be accepted. For 

example: 

a. All of the occupational-related airborne activity originates at Y-12. Y-12 is part 

of the Oak Ridge complex, which includes other sources of radioactive material. 

The inconsistent correlation of monitored airborne uranium concentrations 

elsewhere on the Oak Ridge reservation to releases from Y-12, especially at 

Station 12, calls this hypothesis into question. 

b. In Section 4.2.5, the TBD attributes the poor correlation at Station 12 to the 

paucity of data. The smaller number of data points (8 versus a maximum of 17) 

does not account for the virtual lack of correlation shown in Figure 4.2.5-4. We

believe that fewer data points would lead to a better correlation, both on 

mathematical grounds (fewer degrees of freedom), and on physical grounds: 

meteorological conditions would vary less during a shortened time period. 

Should the Station 12 data not be used in creating a framework for dose 

reconstruction, due to this lack of correlation?

c. Resuspension of uranium deposited on soil, both inside the Y-12 perimeter and 

off site, is not included. Resuspension of deposition during previous years would 

contribute to the uranium concentration in air, yet is uncorrelated to uranium 

releases from Y-12. 

d. The influence of other sources of uranium—releases from sources outside Y-12 or 

resuspension from on-site soils—is indicated by the data for 1993. The uranium

release of 3 kg for that year is the lowest during the period that Stations 2 and 12 

were operating, and the second lowest during the 17-year operating period of 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

137 of 203 

 

Stations 4 and 8. The 1993 calculated χ/Q's for all four stations are the highest of 

any year that these stations operated. At least for the data from Station 12, which 

SC&A subjected to an independent statistical analysis, there is a negative

correlation between the χ/Q's calculated by ORAU and the annual releases of 

uranium: the lower the releases, the higher the χ/Q. 

e. The TBD states that the calculated χ/Q's follow a lognormal distribution with a 

high-end tail. Yet, in assigning 95th percentile values, where the high-end tail 

would have the greatest influence, that aberration from lognormal behavior is not 

addressed. An examination of the data in Tables 4.2.3-1 through Table 4.2.3-4 

reveals the following: 

(i) Station 2: Out of 12 calculated χ/Q's, one is significantly higher than the 95th

percentile value (2.00 × 10-14

 vs. 1.31 × 10-14). Statistically, the highest of 12 

values would correspond to the 92nd percentile. 

(ii) Station 4: Out of 17 calculated χ/Q's, one is over 17 times higher than the 95th

percentile value (5.03 × 10-14

 vs. 8.73 × 10-14). Statistically, the highest of 17 

values would correspond to the 94th percentile. 

(iii)Station 8: Out of 17 calculated χ/Q's, two are higher than the 95th percentile 

value. Statistically, the second highest of 17 values would correspond to the 

88th percentile. 

(iv) Station 12: Out of only eight calculated χ/Q's, one is higher than the 95th

percentile value. Statistically, the highest of eight values would correspond to 

the 88th percentile. 

The above observations indicate that the distributions are not lognormal, and that the 

assumptions used by ORAU to calculate the 95th percentile χ/Q's are not claimant 

favorable, aside from the questions raised earlier in this discussion about the validity of 

the empirical χ/Q approach. This conclusion is further buttressed by our statistical 

analysis of the Station 12 data,1

 which exhibit the best correlation between the annual 

releases and the annual average air concentrations of uranium of the four monitoring 

stations. Subjecting the calculated χ/Q's to the W test

2

 shows that there is a 10% to 50% 

probability that the distribution is lognormal. In other words, there is a better than a 50% 

chance that the distribution is not lognormal. 

6. As a general observation, we question the decision to base the dose reconstruction on the 

calculated χ/Q's, even for years for which measurements of air concentrations exist. If 

the purpose of dose reconstruction is to determine the doses to individual claimants, not 

1

 Our ability to perform numerical analyses on the data in the TBD is hampered by the password protection of 

the PDF files of this and other OCAS documents. This protection prevents us from copying the tables into our own 

software. Allowing SC&A unrestricted access to the contents of these documents would enhance our ability to analyze 

these data.

2

 Richard Gilbert (1987). Statistical Methods for Environmental Pollution Monitoring, Section 12.3.1.
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

138 of 203 

to maintain a uniformity of approach with a bias towards a statistical construct as 

opposed to available real data, it is not clear how NIOSH’s approach in this instance can 

be considered claimant favorable. 

7. The methodology for calculating the air concentrations of the uranium isotopes, which 

are listed in Attachment D to the TBD, is unclear. Our attempts to replicate some of the 

data in Attachment D, using the information in the main body of the TBD, were 

unsuccessful. Furthermore, some of the data in Table 4.2.4-1, which are the basis of the 

data in Attachment D, are themselves suspect. Notably, based on the masses of the 238

U 

releases listed in the fifth column of that table, the 238U activities released in 1994 and 

1995 should be 0.008 and 6.7 × 10-4 Ci, respectively, rather than 0.002 and 0.0021 Ci, as 

listed. Can NIOSH clarify the methodology used in constructing the Attachment D 

tables, and how it verifies the data in Table 4.2.4-1?

8. On p. 26, the TBD states: “. . . it is conservative to assume that the air concentrations 

reported from 1996 to 2002 are equal to the concentrations reported for 1995.” First, 

there are no air concentration data listed for the years 2000 – 2002. The TBD states that 

the data for these years, collected by TDEC, are not usable. More important, this 

statement is contradicted by the data that are listed. The annual average concentrations at 

Station 4 for the years 1996 – 1998 are higher than for 1995. Furthermore, the data in 

Table 4.2.4-1 show that the uranium releases in each year from 1996 through 2001 were 

higher than in 1995. Can NIOSH clarify these seeming inconsistencies? 

9. NIOSH needs to furnish guidance to dose reconstructors for applying the data tabulated 

in Attachment D. The tables list both 50th and 95th percentile values. Presumably, the 

50th percentiles are to be used as the best estimates, and the 5th percentiles as the upper 

bound values. As mentioned earlier, the calculated 95th percentile values are not upper 

bounds, as they are exceeded by annual average χ/Q's at each of the four monitoring 

stations. Can NIOSH clarify why such guidance is not needed?

10. The contributions of 236U are missing in Attachment D. Concentrations of 236U, which 

are listed in Tables 4.2.3-1 through Table 4.2.3-4, are comparable to those of 238U and 

should presumably be taken into account. Can NIOSH clarify why these data are not 

provided?

11. We question the conclusion made in the TBD that the aerial surveys performed by the 

EG&G Remote Sensing Laboratory cannot be used for dose reconstruction. By spanning 

a period from 1973 to 1989, they provide a historical record of the external exposures on 

the site. Such data could be used to provide a temporal trend, and should be combined 

with the 1985 – 1987 scoping survey. The spatial resolution of the aerial surveys is 

comparable to the size of the grid blocks in the scoping survey. Furthermore, the 

exposure rates calculated at 1 m above ground, as presented in the aerial survey reports, 

appear to be more relevant to dose reconstruction than the ground level rates measured in 

the scoping survey. Can NIOSH further justify why these data are not usable for this 

purpose?
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

139 of 203 

12. Given the purpose of the survey and the instruments used, we believe that the exposure 

rates were based on the count rates registered by the sodium iodide crystal in the survey 

meter. Such meters are commonly calibrated with a 137Cs source—the scale on the rate 

meter is set to the calculated exposure rate at the location of the detector. This calibration 

is thus valid only for the principal 137Cs γ-ray and will vary significantly with photons of 

different energies. The meter readings should be converted to dose rates in air (i.e., air 

kerma), using calibration curves that can be obtained from the manufacturer of the meter 

in question, or from manufacturers of similar instruments, and the known spectrum of the 

uranium isotopes deposited on the ground. (See the response curve for a 1” x 1” NaI 

crystal, below). The aerial survey data, which includes some spectral analysis, would be 

useful in this regard. The air kerma should in turn be converted to dose equivalent rates, 

using tables in ICRP Publication 74 for the appropriate exposure geometry. Can NIOSH 

clarify why the above approach would not be justified?

OCCUPATIONAL INTERNAL DOSE (ORAUT-TKBS-0014-5) 

1. The TBD (pg. 24) indicates that methodological problems occurred during the early 

period of uranium bioassay, resulting in an underestimate by a factor of two. How has 

NIOSH/ORAU accounted for this underestimation in the dose reconstructions? 

2. The TBD (pg. 13) discusses process streams of uranium. Included in the process streams 

are what are termed as “solvent extraction raffinates.” What are “solvent extraction 

raffinates”? Is this a process that may concentrate particular radionuclides including 

daughter products?

3. The chemical forms of uranium at Y-12 include S, M and F type materials. What is the 

basis for the use of Type M material from 1948-June 1998 without regard to the type of 

cancer? 

4. Provide an example of how the recycled uranium impurities are accounted for in the dose 

reconstruction. 

5. In the TBD (pg. 23) there is a reference to section 5.4.2 for discussion of missed doses. 

This section appears to be missing. Is it going to be included in the near future? The TBD 

mentions that there is a concern that the technique used before May 1952 could have 

underestimated the urinary uranium concentrations. How is the dose reconstructor going 

to deal with these underestimates of urinary uranium concentrations ? 

6. Was tritium dose included in the whole-body dose at Y-12?

7. Is NIOSH considering issues of frequency of monitoring during the remediation of the S3 ponds (Section 2.6 of the Site Description)?

8. Has NIOSH investigated the incidents or the potential for acute exposures during the 

remediation of the S-3 ponds?
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

140 of 203 

9. How does NIOSH plan to address early period internal dose issues?

10. What were the processes involving Th-232 at Y-12? 

11. Page 7: Is there any explanation for why no urinalysis data have been found prior to 

1948? The fluorometric technique should have been well established by that date. What 

steps are in hand to resolve this issue? Does the discussion of lack of urinalysis prior to 

1948 (p. 5, vol. 5) mean that NIOSH is not doing dose reconstruction for the early period 

at this time? 

12. Page 8: From the earliest days, it was recognized that ‘insoluble’ uranium (except for 

high-fired oxides) exhibited a solubility intermediate between Classes W and Y. This 

was addressed in one way by defining Class Q. However, was any attempt ever made to 

develop a model specifically for these uranium compounds and to then interpret the Y-12 

data explicitly with that model? More generally, this report addresses joint interpretation 

of the urinary and fecal data, but says nothing about interpreting these data in conjunction 

with the chest counting data. Has such an overall analysis been attempted either at the 

level of the individual or in relation to groups of workers with similar exposure histories?

13. Page 9: Figure 5-2 gives a composite urinary excretion curve for 157 individuals. This 

conforms quite well to Class M behavior. However, this is not unexpected at a plant 

where a wide variety of uranium compounds of different degrees of solubility were used. 

It would be much more interesting to see the curves for the individual workers to see if

they span the full range from Class F to Class S (or even beyond). Is this possible? 

14. Page 10: The particle sizes range up to large values (10 μm physical diameter). Are 

there particle size data for respiratory zone relative to general area monitoring and can 

any useful relationships be established between the two? 

15. Page 11: There is some indication in the appendix to the report that distinctions were 

made between routine and special monitoring. It seems likely that special monitoring 

would have been triggered by individual incidents. Could an account be given of the 

special monitoring program employed, including triggers for special monitoring, duration 

of follow up and any analyses of the results that have been undertaken?

16. Page 20: Can the constant 8 in equation 3.1 be reconstructed based on a detailed 

knowledge of the assay procedure used?

17. Page 21: Here and elsewhere a nominal counting efficiency of 0.5 cpm/dpm is used. 

Bearing in mind that this is for counting of a disk, this is the maximum possible value that 

can be achieved. This seems implausible. Can details of the counting geometry be 

provided, so that this issue can be examined?

18. Page 22: For the recent data on uranium in urine by alpha spectrometry, have 

comparisons been made with other methods, e.g. ICPMS or AMS? More generally, how 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

141 of 203 

well are the various assay procedures used verified by inter-laboratory comparison 

exercises?

19. Page 27: Fecal sampling is effective at recording recent intakes of insoluble materials by 

ingestion and inhalation. Has consideration been given to whether fecal samples obtained 

at an interval of 53 days give a meaningful representation even of chronic intakes, which 

will exhibit considerable temporal variability? 

20. Page 35: From what source can Wilcox (1999) be obtained? 

INTERNAL DOSIMETRY CO-WORKER DATA 

1. This report appears to, first, average urinary excretion data over individuals, and then 

interpret the results in terms of Class M and Class S intakes. Can NIOSH clarify why 

this was done, as the results must reflect different numbers of workers in different 

contexts? Also the results will inevitably be smeared in time because of the different 

time courses of exposure of the individual workers, as each worker will have been 

associated with multiple determinations of uranium in urine. Why was this analysis not 

undertaken on an individual-by-individual basis? This would have yielded a much richer 

understanding of the range of exposure conditions. 

2. Given that multiple intake periods were used, the fitting model should be able to give a 

very accurate representation of the data. However, there are significant divergences for 

the Class M result (Figures A-9 and A-10), and these are much worse for the Class S 

analysis (Figures A-19 and A-20). These divergences arise from fitting data for each of

the intervals separately and then adding the results. Why was this done instead of 

undertaking a simultaneous joint fit to all the intervals, or at least rescaling each of the 

individual fits by an overall normalization factor to give a better overall fit? (However, 

this remark should be seen in the context of the previous consideration that an individualbased analysis is to be preferred to the group analysis presented here). 

3. OCAS-TIB-0029, Internal Dosimetry Coworker Data for Y-12 calculates the 50th and 

84th percentiles for urinalysis data. What is the basis for choosing the 84th percentile? 

4. Many of the tables and graphs in OCAS-TIB-0029, Internal Dosimetry Coworker Data 

for Y-12 are missing the time period from 5/1/1952-7/31/1953. Is this intentional or 

should this time period appear in all relevant tables and graphs?

OCCUPATIONAL EXTERNAL DOSE (ORAUT-TKBS-0014-6) 

1. Has NIOSH determined specifically which 25% of the population was monitored from 

1949–1960? 

2. NIOSH/ORAU has developed a photon dose correction factor for those workers who 

wore their dosimeter at their collar when the radiation source was at their waist. This is 

specific to unloading and sorting DU scrap materials, shearing of large pieces of scrap 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

142 of 203 

materials, cleaning of the scrap materials, crucible loading during the melting and casting 

operations, and materials sampling (pg. 43). How are dose reconstructors able to 

distinguish individuals involved in these tasks from other individuals? 

3. Considering the fact that some workers, especially those performing waist-level uranium

handling jobs, could have received beta doses to the skin, breast, and testes, why has 

NIOSH/ORAU opted not to address them? 

4. Elaborate on the statement (pg. 8), “The fabrication of weapon parts was expanded over 

the years to include other radioactive and non-radioactive material.” 

5. Does Table 6.3.4.3-2 (pg. 27) include zero badge results, including those from NTA 

badges not processed? 

6. From the TBD, it is not clear if beta (and low-energy photon) doses were measured and 

entered into the worker’s record and used in dose reconstruction. It states in several 

places that the two-element film badge, and the TLDs, had an open window and an 

effective 1 g/cm2

 filter. But it never states whether the open window dosimeter was read 

and recorded in the dose records. Was the shallow dose read and recorded as a shallow 

dose, Hp(0.07)? Will this information be used in dose reconstruction? 

7. There were likely many accidents and/or incidents (e.g., 1958 Criticality Accident) that 

resulted in unexpected exposures to workers. How extensively has NIOSH/ORAU 

investigated the likelihood of other accidents and/or incidents that may significantly 

impact worker exposures? 

8. Page 16, Section 6.3.2.2 is somewhat confusing. Does the summary below adequately 

characterize the Y-12 neutron monitoring history? 

1943–1950 = No neutron monitoring 

1950–1980 = NTA film

1980–1985 = NTA film for fast neutrons + TLND for other energy neutrons 

1986–1989 = TLND only 

1989–2003 = Albedo-type TLND for proper Hp(10) 

9. In the fourth sentence of Section 6.3.2.2 (Page 16), it states that between 1980 and 1989 

there is a serious gap in the neutron dosimetry information for Y-12. If the neutron 

dosimetry was as listed above, why is this? It appears that the serious gap in neutron 

monitoring occurred from1943–1950. 

10. Page 18, Table 6.3.3.1-1 states in the radiation quantity bias section that the calibration 

sources used may have caused a slight (about 3%) under-response in the recorded dose. 

This is the only place that the figure of 3% appears. How was the value 3% derived, and 

is it used anywhere in the dose reconstruction process? 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

143 of 203 

11. Has NIOSH ruled out the possibility that radiation exposure was higher for Y-12 workers 

exposed to external penetrating radiation from handling recycled uranium (which may 

include other uranium isotopes, such as uranium-232), than for other Y-12 workers in 

other operations?

12. The TBD seems to ignore assignment of external dose for 1943–1950. In the absence of 

records, why has the TBD not developed a methodology for assigning upper bound 

external doses based on field radiological or source term data?

13. Why does the TBD assume that NTA film can effectively record neutron exposure at 

500 keV and above, when the effectiveness of the NTA film falls off significantly 

between 500 and 1000 keV?

14. The PNL neutron spectrum report does not necessarily represent the neutron spectra in 

the field, especially with fission neutrons. What other basis does NIOSH/ORAU have for 

the assumption that 90%–95% of the neutrons in the field are detected by the NTA film? 

15. Why has NIOSH opted to use primarily NTA film results at the Y-12 Plant, yet they have 

specifically excluded the use of these results at other facilities such as Hanford and the 

Savannah River Site?

16. NIOSH/ORAU has prepared ORAUT-RPRT-0032, Historical Evaluation of the Film 

Badge Dosimetry Program at the Y-12 Facility in Oak Ridge, Tennessee: Part 1 – 

Gamma Radiation and ORAUT-RPRT-0033, Historical Evaluation of the Film Badge 

Dosimetry Program at the Y-12 Facility in Oak Ridge, Tennessee: Part 2 – Neutron 

Radiation. How are these reports related to the external dosimetry TBD? Are they 

intended to provide direction to the dose reconstructors?

17. ORAUT-PROC-0042, Accounting for Incomplete Personnel Monitoring Data on 

Penetrating Gamma-Ray Doses to Workers in Radiation Areas at the Oak Ridge Y-12 

Plant Prior to 1961, provides direction on how to assign photon doses prior to 1961. 

What is the purpose for assigning a scaling factor? Explain how the scaling factor is 

applied to the dose reconstruction. 

Document Requests: 

(1) List of Classified Documents Used in the Preparation of the TBD 

(2) Full-scale color representation of the four sets of aerial survey reports 

(3) Reports relating to the scoping survey reference in the Environmental TBD. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

144 of 203 

ATTACHMENT 4: SUMMARY OF CONFERENCE CALLS ON SC&A 

QUESTIONS PROVIDED TO NIOSH 

Dates: June 7, 2005, Medical TBD 

June 7, 2005, External TBD, Part I 

June 8, 2005, Site Description TBD 

June 8, 2005, Internal TBD 

June10, 2005, Environmental TBD (Provided in Attachment 5) 

June13, 2005, External, TBD, Part II (Combined with Part 1 Above) 

Y-12 Occupational Medical Dose TBD Conference Call Summary 

Date: June 7, 2005 

Subject: Occupational Medical Dose TBD, ORAUT-TKBS-0014-3 Rev. 00 PC-1, 

 (Murray 2004) 

Time: 9:00 to 9:45 a.m.

Individuals who participated included: 

ORAU: William Murray, Elyse Thomas, and Ed Scalsky 

NIOSH: Tim Taulbee and Greg DeCecco 

SC&A: John Mauro, Joe Fitzgerald, Tom Bell, Hans Behling, and Ron Buchanan 

SC&A: How much is really known about the number of photofluorographic (PFG) films that 

were taken in the 1943 to 1947 timeframe. Steve Wiley in his April 10, 2002 summary letter 

stated “...to the best of our knowledge, there is no evidence of any fluoroscopic chest x-ray 

examinations performed at the Y-12 Medical x-ray Department”(Wiley 2002). The TBD, 

however, in Section 3.3, page 6 uses the Graham 1946 letter that documents the order of 600 

Eastman single coated x-ray films from Oak Ridge Hospital to come to the conclusion that: 

“Thus, it is clear that pre-employment chest x-rays were taken with a PFG unit from 1943 to 

1947 as evidenced by 4” x 10” films found in the medical records and purchasing records.” 

What do you know about the number of PFG chest films taken from 1943 to 1947 and what 

groups received them?

ORAU: We don’t know a lot more than what you have stated. We did find evidence in the 

individual medical records of the presence of 4” by 10” chest films, which makes it seem clear 

that they did take PFG films. We have documented that a PFG unit was sent to Y-12 by our find 

of a telegram that Y-12 sent to General Electric in Nashville about the set up of the PFG unit. 

We also have the information that 6000 films were ordered from Eastman as you stated from the 

Graham 1946 request. We also have found out that there were several groups on which PFG 

chest films were more commonly used: 

 Food Handlers 

 Cafeteria Workers 

 Uranium workers 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

145 of 203 

SC&A: For those workers who have PFG chest x-rays multiple times a year, did their medical 

records record each PFG chest x-ray examination? 

ORAU: We also have evidence from their medical records that these groups often had several 

of these PFG films done each year. There would often be 3-4 PFG x-rays, which seems to 

capture the number of times the PFG x-rays were taken. 

SC&A: Does this mean that the frequencies of occupational posterior/anterior chest x-rays at

Y-12 in Table 3C-1, page 15 of the occupational medical dose TBD based on Wiley 2002, could 

be greater so that some workers got more frequent chest x-rays?

ORAU: In the case of those special groups who got multiple PFG chest x-rays, the frequency 

would be greater. It is possible that this might be true for some others but the individual medical 

record is likely to document them. 

SC&A: SC&A has a concern about how accurate the 3.0 cGy (3.0 rem) entrance kerma default 

dose for photofluorographic chest x-rays are. How well documented is this entrance kerma dose 

of 3cGy for PFG x-rays? If this is off by a lot, it could mean a big difference in the skin entrance 

kerma dose. I realize this is addressed in ORAUT-OTIB-0006 (Kathren 2003), in Section 3.3 

and Table 3.3-1, page 17, but I wondered what you have found on this subject. 

ORAU: You are correct, the most we know is what is in OTIB-0006 and Ron Kathren is very 

credible. We had access to what Ron Kathren was preparing as we were finalizing the Y-12 

occupational medical dose TBD. 

SC&A: It is noted in ORAUT-OTIB-0006, Rev. 02, page 17, (Kathren 2003) that additional 

information is available about Table 3-4, page 8 of the occupational medical dose TBD. Since 

this was available as the occupational medical dose TBD (ORAUT-TKBS-0014-3, Rev. 00 PC1, 2004) was being finalized, would it not have been beneficial to provide this information in the 

occupational medical dose TBD?

ORAU: Since we knew the OTIB-0002 would follow our occupational medical dose TBD 

shortly, we decided to let the OTIB-0002 provide the details. 

SC&A: It is noted in the occupational medical dose TBD stated “The medical practices used at 

Y-12 are assumed to have followed the adoption of standards of radiology practice during the 

1930s and 1949s to minimize dose to the patient. However, there is the potential for significant 

dose from occupational medical x-ray examinations, depending on the type of equipment, the 

technique factors, the number of photofluouographic (PFG) examinations typical in the early 

years, and the number of radiographic examination (Cardarelli et al. 2002).” On what basis did 

you make that assumption?

ORAU: It was the information provided in the Cardarelli 2002 reference cite that led us to make 

that assumption. We also spoke with an x-ray medical technologist that worked at Y-12 and also 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

146 of 203 

collaborative this assumption. He was there in the early period, left for a while and returned in 

1971. We, however, have no information on the calibration of these early chest x-ray units. 

SC&A: On page 9 of the occupational medical dose TBD, you have stated that: “Other 

variables, such as the use of screens and grids, reciprocity failure, film speed and development 

would not affect the beam output intensity. On what basis did you make that assumption?

ORAU: Although we did not go into detail on this point in the occupational medical dose TBD, 

Ron Kathren in OTIB-0006 does provide the rationale for this assumption. 

SC&A: It is noted in the occupational medical dose TBD, page 7, Table 3-2, that the description 

of the Type II machine used from 1948 to 1968 that the Type II equipment is unknown. 

However, in Table 3-3 on the same page specific technique factor are provided for each type of 

x-ray equipment included the Type II machine. How can you be certain that these technique 

factors for the Type II machine in Table3-3 are correct? Wiley 2002 also mentions that the exact 

model of the medical diagnostic x-ray machine was unknown and that the filtration was 

unknown. This would imply that knowing anything about measurement output is also lacking. 

ORAU: We know little more that what was provided in Wiley 2002. The x-ray technologists 

advised us that the unit used was similar to standard units in use at that time and provided us 

information on the Type II x-ray machine at Y-12. As we stated in Section 3.4, page 7 of the 

occupational medical dose TBD, in the period from 1943 to 1947, no actual x-ray output 

measurements are available. 

SC&A: We note in Wiley 2002, that he states that radiation exposure from the early chest x-ray 

units was 30 mr for the period between mid 1940s to early 1960s. It drops to a radiation 

exposure of 20 mr between the early 1960s and January 1982 and after that was further reduced 

to 10 mr. How do these radiation exposure values relate to the entrance kerma values in Table 

3.4, page 8 of the occupational medical dose TBD? For instance that table shows the entrance 

kerma in cGy for the pre-1970 period for a posterior/anterior Y-12 chest x-ray as 0.2 cGy (or 0.2 

rem). 

ORAU: We believe that the 30 mrem in the early period is the dose measurement in air. That is 

then converted into a skin entrance kerma dose by methods described in the NCRP Report No. 

102 (NCRP 1989) cited in footnote "a" of Table 3-4. 

SC&A: How were the organ doses from photofluorographic chest x-rays (rem) in Table 3C-2, 

page 17, of the occupational medical dose TBD determined? Did you use the entrance kerma of 

3.0 rem and multiply it my some factor provided in ICRP 34 (1992)? I bring this up because if 

there are questions about how accurate these organ doses are, then a claimant may have gotten 

more organ dose that these might indicate. 

ORAU: Our understanding is that ICRP 34 (1992) actually provided the organ doses for each of 

the organs listed in Table 3C-2. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

147 of 203 

SC&A: We note in Table 3C-2 that there is an organ category called “Remainder”. What 

organs fall into that category?

ORAU: “Remainder” represents any other organ that is not available in the list of organs. This 

is addressed by ICRP 34 (1992) and is catch all if you can’t find the organ you are considering. 

SC&A: It is interesting to note that the organ “skin” in Table 3C-2 has the highest organ dose. 

Is the PA dose to the front of the body or the skin of the back: Also would prostate fall under 

Remainder”?

ORAU: Yes skin does have the highest organ dose to be claimant favorable. The doses in Table 

3C-2 is an exit dose so that it is the dose to the skin of the back. Prostate cancer falls under the 

organ “Urinary bladder” 

SC&A: If you are dealing with a skin cancer on the face, do you then use these exit doses at the 

posterior surface of the back. 

ORAU: In the case of skin cancer to the face, we would modify to dose estimate to take into 

account the dose to the skin of the face. 

SC&A: Is there a reason why Table 3.1-2, page 16 in OTIB-0006 was not utilized in the 

occupational medical dose TBD. This table has wealth of summary data on actual beam

measurements for that might be applicable to Y-12 as well. 

ORAU: Since this Table 3.1-2 is actually for the Hanford site, we found that some of it was not 

applicable to Y-12 and this did not include it in the occupational medical dose TBD. 

SC&A: Can you tell us more about the medical impacts of the June 1958 Y-12 criticality 

incident. 

NIOSH: There were eight workers who were involved in the Y-12 criticality accident. They 

were all followed closely at the time and their medical records provide a lot of documentation. 

The dosimetry was quite well done. These have been well written up in the literature and a 

search on the “O” drive under “criticality” should bring up the pertinent documents. Six of these 

workers have since filed claims and all six have been awarded. 

SC&A: Did these workers later develop cancer? 

ORAU: Yes, all six got cancer. 

SC&A: Since you have no data on beam output, how well do you think you can characterize 

entrance skin kerma for the older Type I and Type II x-ray machines? 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

148 of 203 

ORAU: Since there was no collimation during these early periods, we have assumed the 

maximum beam size was delivered which is the most claimant favorable assumption we can 

make. 

Y-12 Occupational External Dosimetry TBD Conference Call Summary

Date: June 7, 2005 

Subject: Occupational External Dosimetry TBD, ORAUT-TKBS-0014-6, Rev. 00, (Kerr 2003) 

Time: 9:00-10:30 am EDT 

Individuals who participated included: 

 

ORAU: Bill Murray, and Elyse Thomas 

NIOSH: Tim Taulbee and Greg DeCecco 

SC&A: John Mauro, Ron Buchanan, and Bob Alvarez 

Date: June 13, 2005 

Time: 10:00-11:00 am EDT 

Individuals who participated included: 

ORAU: George Kerr, Ed Scalsky, Bill Murray

NIOSH: Stu Hinnefeld 

SC&A: Ron Buchanan, Tom Bell, and Kathy Robertson-DeMers 

ORAU General Comments: 

We will mainly concentrate on the Y-12 TBDs related to external dosimetry, and not on dose 

reconstruction or internal dose issues. 

SC&A: Has NIOSH determined specifically which 25% of the population was monitored from

1949–1960? 

ORAU: Workers that had the potential to be exposed to radiation in the Assay Labs, 

Radiographic Shop, Spectrographic Shop, and the Metal Machine Shops., as well as any X-ray 

workers and radioisotope handlers were badged. 

SC&A: NIOSH/ORAU has developed a photon dose correction factor for those workers who 

wore their dosimeter at their collar when the radiation source was at their waist. This is specific 

to unloading and sorting DU scrap materials, shearing of large pieces of scrap materials, cleaning 

of the scrap materials, crucible loading during the melting and casting operations, and materials 

sampling (pg. 43). How are dose reconstructors able to distinguish individuals involved in these 

tasks from other individuals? 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

149 of 203 

ORAU: These workers are identified through interviews and job changes in their personnel 

files. If they performed that kind of task, the 1.34 x correction factor will be applied to their 

recorded dose. 

SC&A: Considering the fact that some workers, especially those performing waist-level 

uranium handling jobs, could have received beta doses to the skin, breast, and testes, why has 

NIOSH/ORAU opted not to address them? 

ORAU: While this version of the TBD does not cover this subject, a revised TBD will be issued 

that includes this dose reconstruction (DR). 

SC&A: Elaborate on the statement (pg. 8), “The fabrication of weapon parts was expanded over 

the years to include other radioactive and non-radioactive material.” 

ORAU: The unclassified information is contained in Pages 6-8 of TKBS-0014-02. Some other 

information will come out in the revised TBDs/TIBs concerning deuterium, thorium, and beta 

dosimetry; perhaps by fall of 2005. The history of Y-12 by Wilcox might also provide some

additional information. 

SC&A: Does Table 6.3.4.3-2 (pg. 27) include zero badge results, including those from TLND 

(not NTA) badges not processed?

ORAU: Yes. It includes all issued neutron badges. All neutron badges during this time period 

were read. It would also include zero or MDL reading also. The MDL was small, in the order of 

a few millirem. 

SC&A: From the TBD, it is not clear if beta (and low-energy photon) doses were measured and 

entered into the worker’s record and used in dose reconstruction. It states in several places that 

the two-element film badge, and the TLDs, had an open window and an effective 1 g/cm2

 filter. 

But it never states whether the open window dosimeter was read and recorded in the dose 

records. Was the shallow dose read and recorded as a shallow dose, Hp(0.07)? Will this 

information be used in dose reconstruction? 

ORAU: Yes, the open window and filtered film was read, recorded, and will be used in the DR 

process. The penetrating Hp(10) dose includes the gammas plus neutron dose and the Hp(0.07) 

includes the penetrating plus the skin dose. Most of the beta dose was received by the metal 

workers. A future TIB will be issued that will address the beta and low energy photon dose in 

further detail. 

SC&A: There were likely many accidents and/or incidents (e.g., 1958 Criticality Accident) that 

resulted in unexpected exposures to workers. How extensively has NIOSH/ORAU investigated 

the likelihood of other accidents and/or incidents that may significantly impact worker 

exposures? 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

150 of 203 

ORAU: NIOSH’s instructions were not to include accidents/incidents in the TBDs, unless they 

involved a significant number of workers, such as the 1958 criticality accident. The DR can 

obtain information concerning this issue in the individual worker’s files, or interviews. 

SC&A: Page 16, Section 6.3.2.2, is somewhat confusing. Does the summary below adequately 

characterize the Y-12 neutron monitoring history? 

1943–1950 ― No neutron monitoring. 

1950–1980 ― NTA film. 

1980–1985 ― NTA film for fast neutrons + TLND for other energy neutrons. 

1986–1989 ― TLND only. 

1989–2003 ― Albedo-type TLND for proper Hp(10). 

In the fourth sentence of Section 6.3.2.2 (Page 16), it states that between 1980 and 1989 there is 

a serious gap in the neutron dosimetry information for Y-12. If the neutron dosimetry was as 

listed above, why is this? It appears that the serious gap in neutron monitoring occurred 

from1943–1950. 

ORAU: NTA film for neutron monitor was most likely used on a routine basis around 1950. 

Most of the neutron dose at Y-12 was because of the operation of the 86” cyclotron between 

1950 and 1961. During 1962–1980 the ORNL operation of this unit did not involve Y-12 

workers. Neutron doses were so small during 1980–1989 that ORNL processed the neutron 

dosimeters for Y-12. Therefore, this processing information would be at ORNL. There is not 

really a problem with the neutron dose, but Y-12 records do not contain the details because of the 

ORNL processing. 

SC&A: About how many Y-12 workers received neutron exposures?

ORAU: Until ORAU reviewed the Y-12 epidemiology studies, there was a lot of uncertainty on

this. It appears that up to 1980, there were 375 positive neutron recorded doses for 

approximately 143 workers. Some of these were assigned the MDL of 50 mrem. The dose was 

often less that the MLD but the MDL dose of 50 mrem was assigned anyway. From 1962 to 

1980 there were no neutron doses or is was so low that it couldn’t be detected by the NTA film. 

Then around 1980, Y-12 switched over to the neutron albedo dosimeter. There was some

medical isotope production done in the early 1980s, but this was done by X-10 workers. There 

were so few Y-12 workers involved that they sent all the neutron dosimeters to X-10 for 

processing. During 1989 and 1990, dosimeters were either read at ORNL or X-10. 

SC&A: The TBD seems to ignore assignment of external dose for 1943–1950. What is being 

done to develop a methodology for assigning upper bound external dose based on field 

radiological and source term data, particularly for neutrons?

ORAU: We have found little monitoring data for 1943 to 1950 and it appears that there was not 

much of a neutron problem during this period. With the (n/alpaha) reaction in UF4 it is doubtful 

that you could measure the neutron fluence. I think they used the Ying and Yang dual ionization 

chamber. This was written up by Hurst and Richie in the late 1940s. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

151 of 203 

SC&A: It appears the external film badge was not perfected until about 1948. What did they 

use for monitoring prior to that?

ORAU: At Y-12, the started with the Victoreen ionization chamber. In 1940, they switched to 

the use of the film badge. During 1948–1949, Y-12 used both the film badge and the pocket 

ionization chambers. An extrapolation process was used where needed. A report providing 

more details on this is in preparation for NIOSH review. 

SC&A: Then what you are saying is that there will not be many dosimetry records prior to 

1950. Is that correct? 

ORAU: During the Tennessee Eastman Company (TEC) period, most of the concern was about 

x-rays from high voltage tubes in the calibration panels and the rectifiers (high voltage vacuum

tubes.) During this period, the pocket ionization chambers were the only area monitors. They 

used the Victoreen R chambers (pocket dosimeters) to ensure they were minimizing x-ray dose 

from the Calutron high voltage panels. 

SC&A: Page 18, Table 6.3.3.1-1 states in the radiation quantity bias section that the calibration 

sources used may have caused a slight (about 3%) under-response in the recorded dose. This is 

the only place that the figure of 3% appears. How was the value 3% derived, and is it used 

anywhere in the dose reconstruction process? 

ORAU: This value was taken from NBS-85 (1962). This 3% correction factor is so small that it 

is likely that it would not be applied by the DR as it is within the noise range of the calibration 

and film badge reading process. The uncertainty of calibrating and reading the film badge is 

about 10% and the 3% is just a part of that. 

SC&A: Has NIOSH ruled out the possibility that radiation exposure was higher for Y-12 

workers exposed to external penetrating radiation from handling recycled uranium (which may 

include other uranium isotopes, such as uranium-232), than for other Y-12 workers in other 

operations?

ORAU: The RU workers were monitored for radiation exposures. We are currently working on 

addressing the beta doses, either from first-principles or the opened literature. 

SC&A: The TBD seems to ignore assignment of external dose for 1943–1950. In the absence 

of records, why has the TBD not developed a methodology for assigning upper bound external 

doses based on field radiological or source term data?

ORAU: The records to date do not indicate any significant neutron sources before 1950 when 

NTA film was first used. Photon exposures were monitored using “R” chambers, gamma-film

badges, and pocket ionizations chambers. 

SC&A: Why does the TBD assume that NTA film can effectively record neutron exposure at 

500 keV and above, when the effectiveness of the NTA film falls off significantly between 500 

and 1000 keV?
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

152 of 203 

ORAU: This was the threshold chosen for Y-12. If the threshold was raised to 800 keV, it 

probably would not make much difference in total neutron dose because most of the neutrons 

were not of the lower energy fission spectrum at Y-12. 

SC&A: The PNL neutron spectrum report does not necessarily represent the neutron spectra in 

the field, especially with fission neutrons. What other basis does NIOSH/ORAU have for the 

assumption that 90-95% of the neutrons in the field are detected by the NTA film? 

ORAU: There were very few fission neutrons at Y-12. The critical assembly was most likely 

the only significant source because (alpha,n) produced neutron exposures were extremely low in 

the work environment. The major source of neutrons, besides the 86” cyclotron, was the 

radioisotope sources, such as RaBe, PuBe, and AmBe; the PNL measurements match these 

sources. Most of the sources were in shielding facilities. Most calibration sources were kept in 

pigs and only brought up by remote control. Therefore, there is likely not much exposure from

these radioisotope sources. The Y-12 neutron report covers some of this dating back to 1949 to 

the early 1950s. 

SC&A: Why has NIOSH opted to use primarily NTA film results at the Y-12 Plant, yet they 

have specifically excluded the use of these results at other facilities such as Hanford and the 

Savannah River Site?

ORAU: Most of the neutron energies at Y-12 were above the NTA film threshold. Therefore, 

the NTA film is more applicable at Y-12 as compared to other AEC/DOE sites that had more 

fission neutron spectra. The use of n/p ratio measurements to estimate neutron doses is 

addressed further in the ORAUT RPRT-0033 document, which is available for use by the DRs 

when needed. Usually if you don’t see a positive neutron dose then you can assume that workers 

were not exposed to neutrons. 

SC&A: NIOSH/ORAU has prepared ORAUT-RPRT-0032, Historical Evaluation of the Film 

Badge Dosimetry Program at the Y-12 Facility in Oak Ridge, Tennessee: Part 1 – Gamma 

Radiation and ORAUT-RPRT-0033, Historical Evaluation of the Film Badge Dosimetry 

Program at the Y-12 Facility in Oak Ridge, Tennessee: Part 2 – Neutron Radiation. How are 

these reports related to the external dosimetry TBD? Are they intended to provide direction to 

the dose reconstructors? 

ORAU: Yes, they are available to the DRs and can be used to supplement the TBDs. 

SC&A: It is stated on page 7 of the external dosimetry TBD that the fabrication of weapon parts 

was expanded over the years to include other radioactive and non-radioactive materials. What 

are some of these other radionuclides?

ORAU: Y-12 process lithium, DU, 233U and thorium. We have been looking at the potential for 

beta dose and are trying to develop methods to deal with these other radionuclides. A report by 

William Wilcox published in 1999 provides an overview of the history of Y-12 and may have 

some information that can help address this. Hand doses are covered in the internal dose TBD. 

In Rev. 01 of the internal dose TBD, there is a lot of data on thorium air sampling and whole 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

153 of 203 

body counting conducted when workers were suspected of being exposed to thorium. In over 

100,000 whole body counts taken, only 160 were found to be positive. 400 to 500 have been 

found to be below the MDL of 6 mg for thorium. A thorium TIB is being reviewed by OCAS 

that will, when published, provide a lot of new information. Some of the data on this is still 

“Official Use Only” but what is unclassified will be made available to the public. 

SC&A: The external dosimetry TBD, page 8, mentions that for the later program involved in 

the fabrication of weapons parts, that only about 25% of the workers were monitored up to the 

time of the criticality accident at Y-12 in 1958. What were the groups that were monitored?

ORAU: The workers that were monitored were those in the SA Lab, the metal works facility, 

the spectrographic metal shop, the metal machine shops and the assay labs. It has been hard to 

find specific data. Much of it is in documents which contain classified as well as some

unclassified data and cannot be easily used prior to be cleared for release by the declassifiers. 

What we are seeing is that if area monitors indicated that workers could exceed 10% of the 

guidelines at the time, these workers were put on monitoring programs. If they changed jobs and 

no longer in jobs that might result in them getting over 10% of the guidelines, they were taken 

off the monitoring program or added if they were found to move to jobs where they might 

exceed 10% of the guidelines. These workers were tracked and we find evidence of this is the 

procedures at Y-12. Monthly reports and health physics reports often provide these kinds of 

details. We have found that a 1962 summary report has quarterly gamma doses recorded which 

also includes beta skin doses. These seem to be pretty well kept up in following these workers. 

In more recent years, at least from experience from X-10, superiors had to approve work permits 

that involved workers who went into area where the 10% of the dose guidelines might be 

exceeded. Health physics staff tracked workers on a day to day basis. They used pocket 

dosimeters and if found a worker exceeded 0.5 rem, they would pull the worker’s film badge and 

have it developed right away. In more recent years, health physics supervisors are under a 

mandate to ensure no workers do no exceed the 2 rem/year occupation exposure limit set by 

DOE regulations. In general, workers have been well tracked. 

SC&A: Is there anything new on the Y-12 SEC petitions?

NIOSH: There are three SEC petitions for Y-12. There is some information on the NIOSH 

website about these, and it is possible that two of these are already or may soon be posed on the 

website. 

SC&A: ORAUT-PROC-0042, Accounting for Incomplete Personnel Monitoring Data on 

Penetrating Gamma-Ray Doses to Workers in Radiation Areas at the Oak Ridge Y-12 Plant 

Prior to 1961, provides direction on how to assign photon doses prior to 1961. What is the 

purpose for assigning a scaling factor? Explain how the scaling factor is applied to the dose 

reconstruction. 

ORAU: The purpose of the scaling factors is to allow the DR to scale up the calculated pre1961dose for a worker that had higher-than-average dose data in the post 1961 period. This will 

takes in consideration that the worker most likely worked in a higher-than-average radiation 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

154 of 203 

environment before 1961 and therefore should be assigned a higher than average dose for the 

pre-1961 period. 

Y-12 Occupational Site Description TBD Conference Call Summary 

Date: June 8, 2005 

Subject: Site Description TBD, ORAUT-TKBS-0014-2 Rev. 01, (Jessin 2005) 

Time: 9:00 to 10:30 a.m.

Individuals who participated included: 

ORAU: William Murray, Karen Jessen, and Ed Scalsky 

NIOSH: Stu Hinnefeld 

SC&A: Joe Fitzgerald, Hans Behling, Joyce Lipsztein, Bob Alvarez, Kathy Robertson-DeMers, 

Tom Bell, and Ron Buchanan 

SC&A: There is much about the history at Hanford that is not thoroughly covered in the Site 

Description TBD. A Health and Human Services Document published in December, 1996 

provides a very detailed review of operations at Y-12 up through October 1996. This provides a 

very detailed look at the operations conducted at Y-12. It appears that NIOSH has not 

thoroughly reviewed and used this data. There were many operational changes that occurred 

within each facility and across the Y-12 site that can significant impact the dose reconstruction 

process. The process histories provided in the HEW, 1996 document provides an important 

chronology of what happened in different buildings and with different processes at the Y-12 site. 

ORAU: The Y-12 documents were the third ones prepared and came early in the development 

of the site specific TBDs. Rev. 00, which came out in March 2003, was supposed to be a general 

discussion of Y-12 activities and was not supposed to include all the data or to have the kind of 

detail you describe. 

SC&A: The operations and activities at the Y-12 site were a dynamic process. New processes 

were being added and some dropped as the site evolved. There was a mixing of workers around 

the Y-12 site, between different facilities on the Y-12 site and even to different area with a 

specific building or facility. There is a need to link processes with the specific records for the 

operation, building or facility. There is often, especially in the early period when there was no 

bioassay for each type of operation. To do an adequate job of reconstructing dose, the DR needs 

to know more about what facility the claimant is working at and the process in which he or she is 

involved. It is also important to refer to any databank that might be available to account for 

accidents, fires, explosions or off-normal incidences that would affect the claimant’s dose. 

ORAU: We all wish there was more time to help develop this clearer picture on the type of 

process and what the individual claimant likely encounter, but the accelerating schedule to get 

these TBDs out for us often precludes getting into this level of detail. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

155 of 203 

SC&A: We realize that, due to fact that some issues are still classified, it is best to use the 

information that is in the public domain. SC&A has an unclassified CDROM that contains 

information on work at Building 99-102. If you provide a point of contact, SC&A will be glad to 

share a copy of the CDROM with you. 

ORAU: We would be most interested in getting a copy. Let’s work out the details after this 

call. 

SC&A: Regarding question 4 in our list of site description TBD questions sent to NIOSH on 

May 25, 2005, we note the TBD mentions the handling of recycled uranium (RU) at Y-12, but 

does not provide any estimates of dose or the timelines when RU was handled. A Task Force 

was convened in 1985 to review the use of RU at Fernald, but the Task Force, looked at Y-12 as 

well. Their report, not the report in 2000 that dealt with mass balances, addressed this review. 

Y-12 has also done its own assessment which explains the handling of RU by workers and which 

groups of workers did not handle RU at Y-12. It seems the highest dose to workers were gamma 

exposures. Y-12 and Savannah River Site (SRS) had a handshake agreement on what the 

specifications were but we have not been able to locate any written information on these 

specifications and what trace contaminants were present. There is also information on plutonium

work at Y-12 in the PU Out of Specification (POOS) document. Such specifications are 

important to locate. As you might know, Fernald had to shut down their RU line because they 

had problems with specifications and it was found that they had more RU there than the 

specifications indicated were possible. At Y-12 and SRS, there are no formal specification 

standards, only these gentlemen’s agreements. Since they were not measuring for trace 

contaminants, there is a potential for missed dose from these trace contaminants of RU. 

ORAU: It was not possible to get into this kind of detail as we prepared the Site Description 

TBD. Table 2.4-1, does provide information on the process chronology of operations that we 

found. 

SC&A: The CDROM we indicated we are willing to send to you should help to expand upon 

Table 2.4-1. It is important to know the relationship on what was being handled, when it 

occurred and at what facilities. Y-12 was handling 233U because we know that Hanford had sent 

some 200 Kg. to Y-12. There was keen civilian and military interest at the time on how the 233U 

was to be used. Unclassified specifications have been found that state that this was only 8 Kg. It 

is also known that Rocky Flats was using 233

U. It looks like at Y-12 in the 1960s that tons of 

233U were processed at Y-12 that was in nitrate solution. ORNL might have a handle on this. 

We have noted that at Hanford they tried to control this to < 15 ppm.

ORAU: We have found no information on the use of 233U when the data was retrieved from

Y-12. Much of the data on 233U is still classified. We only focused on records provided by 

Y-12. 

SC&A: We also understand that recent efforts by Mel Chew has uncovered a number of 

documents that will better define the use of thorium at Y-12 and that this was a major focus of 

Mel Chew’s visit to Y-12 recently. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

156 of 203 

ORAU: Yes, that is correct. 

SC&A: In the review that Mel Chew did, we hear that he focused on 232

Th and 228Th, on whole 

body counting and air sampling. 

ORAU: Yes, that is correct. The documents that Mel Chew collected are now on the ORAU 

database “O” drive. 

SC&A: Oak Ridge and Y-12 workers seemed to be working back and forth between these 

facilities, which make it hard to track who was responsible for their dosimetry. There were also 

some Oak Ridge National Laboratory (ORNL) facilities at the Y-12 site. Some of the analyses 

done on Y-12 workers were ORNL analyses. We wonder, do these fall under X-10 dosimetry or 

Y-12 dosimetry?

ORAU: There were 3 or 4 facilities to which ORNL workers and staff had access to at the Y-12 

site. Some of the ORNL health physics staff did help on Y-12 analyses. 

SC&A: Did Y-12 health physics staff have radiation safety overview or did the X-10 health 

physics staff have this overview?

ORAU: ORNL work at X-10 is covered in the ORNL Site Profile TBD. 

SC&A: Y-12 maintenance personnel were known to show up at the X-10 projects. Y-12 

maintenance personnel went to K-12 and X-10 personnel went to K-12 as well. 

ORAU: Yes these did occur. We have to rely on claimant’s personal file and the CADI to 

determine where and when a claimant worked at these different facilities.

SC&A: A maintenance employee at Y-12 told us that ORNL called him to work in the Biology 

Building at ORNL. How is this handled?

ORAU: We assume that this would come up in the CATI interview with the claimant. For the 

most part, work at Y-12 was covered by Y-12 health physics personnel. 

SC&A: It appears that if a Y-12 employee was on loan to K-25, that his dosimetry would still 

be considered Y-12 work. Is that correct?

ORAU: We don’t see this level of detail, because we do not handle these aspects here at Y-12. 

SC&A: We can assume then that an X-10 worker is subject to radiation safety controls from X10 health physicists?

ORAU: That’s what we understand. 

SC&A: Pertaining to question 7, Table 2.3-1 shows some “NA” notation for some of the 

buildings listed. Does that mean that the actual building number is not available or that it is not 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

157 of 203 

listed for security reasons? Although a footnote “a” is shown in the table, footnote “a” is not 

discussed below the table. 

ORAU: It just means that it is not available. 

SC&A: In regard to question 5, it seems like more information is needed to determine if

radiation exposure data are accurate and complete for people who cycled in and out of the Y-12 

Laboratory and ORNL during the period between 1943 and the early 1970s. The building 

numbers, i.e., the “not available” in Attachment B, Table B.1-1 make it difficult for the dose 

reconstructor to handle determine radiation exposure data. 

ORAU: Yes, you are correct. This kind of cycling between buildings occurred even into the 

late 1980s and early 1990s. 

SC&A: It seems it will be important to get a better feel for this. 

ORAU: We hope to be able to do this. 

SC&A: A Department of Energy Defense Program survey done might be a way to do this. Is 

this report still classified? This included a highly enriched uranium (HU) vulnerability

assessment. This should help to give you some idea of the extent of the problem. The TBD 

would be strengthen if it included this historical information and postulated on the dose 

reconstructor will need to do to develop a dose for the worst case situation. 

ORAU: We did not have this document at the time we wrote the Rev. 01 Site Description TBD. 

It appears that SC&A sees and has access to documents that NIOSH does not have. We have 

been hampered by the shutdown of the Y-12 library which has made it difficult to find Y-12 

documents. We don’t know where they sent their documents when they shut down the Y-12 

library. Some of them have been found at the ORNL library, but only some of them are there. 

George Kerr might know something about where these documents were sent and perhaps you 

can ask him this on you conference call next Monday, June 13, 2005. 

SC&A: It is possible what was once publicly available documents may no longer be available 

because they may have been subsequently classified. 

ORAU: We went to the site and did exhaustive research is what was there. We think we have 

recovered a lot of the important documents. It’s hard to tell if some of these were later classified.

SC&A: ORNL tried to use state-of-the-art measuring devices and dosimetry, but in the early 

days they were operating out of the old wooden buildings which made efforts to clean up 

contamination difficult. There are some reports on this. Have any of these Y-12 reports been 

declassified? 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

158 of 203 

ORAU: No, I think many are still classified. The only possible help might be the data on 

thorium in the occupational internal dose TBD. It is likely that many of these documents may be 

“Official Use Only” and not actually classified higher than this. 

SC&A: In regard to the electromagnetic isotope separation (Calutron) plant operated in the first 

era from 1942 to 1947, were you able to find any dosimetry data or to document to potential for 

dose from uranium contamination?

ORAU: No. 

SC&A: Can you tell us more about what is involved in the Y-12 Special Exposure Cohort 

(SEC) petitions. We understand that Lew Wade may be able to provide some background data 

on them but we have not seen yet the actual SEC Petitions? 

NIOSH: There is background information on the three Y-12 SEC Petitions in a June 6, 2005 

Federal Register Notice. Two of these three SEC Petitions are under review by NIOSH and I 

believe I just saw them posted on the ORAU web site. 

SC&A: In regard to question 10, on whether 232

Th fuel for the Molten Salt reactor or any other 

reactor at Y-12 was ever fabricated there, is it likely the Mel Chew review will address this. It

appears that Mel Chew may be the only person who will be able to address this. 

ORAU: Mel Chew, with the recent records he has just found at Y-12, would be the best person 

to answer that. 

SC&A: In reviewing the HU vulnerability assessment study, we note that the 228

Th has some

hard gamma. Thus exposure to these hard gammas at Y-12 could be important for workers 

doing thorium processing evolutions. It is also unclear whether or not, workers were exposed to 

232U, in particular the dose from its daughter products. The new Mel Chew data on thorium will 

be important. SC&A has provided some useful information on 232U in our soon to be released 

Hanford Site Profile Review Draft Report. 

ORAU: George Kerr, is our subsequent conference call, can perhaps comment on the internal 

dose considerations of thorium and 233U as well. 

SC&A: In regard to question 11, have you located a databank for incidents at Y-12? The only 

one we have found was established in 1990. The Y-12 Vulnerability Assessment refers to such a 

listing of incidents in the 1970s and 1980s but we are not sure any existed prior to that. 

ORAU: We only have information on incidents or accidents that impact a large number of 

individuals. The criticality accident in 1958 is mentioned in the Site Description TBD. We have 

not covered many specific such incidents unless they are major incidents like the criticality 

accident. We presume a worker’s involvement in such an incident or accident will be identified 

in the CATI interview or in the individual claimant’s personal file. 

SC&A: What is considered “large numbers of individuals”?
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

159 of 203 

ORAU: We handled these when we first wrote our TBDs on an individual basis and only 

address major incidents in the TBD. Now we have been adding in more data in the TBDs 

especially when there are large numbers of people involved, it’s a large area, or there are 

significant doses involved in the incident. 

SC&A: In the Y-12 vulnerability assessment document, there were situations that RadCon felt 

were significant. How has NIOSH made adjustments to the dose reconstruction process as a 

consequence of contamination spread listed? There are instances reported where contamination 

was carried around without first going through a decontamination process. The whole west side 

of Y-12 was totally contamination with uranium and various groups were coming and going into 

these areas. How does NIOSH adjust doses to take this into account? As noted in our issue 13, 

this applies to other uranium chemicals as well. 

ORAU: The Calutron Plant did not operate too long. It was shut down in 1947 and a new 

gaseous diffusion plant replaced it. After 1947, additional health physics resources were put in 

place to address controls for contamination spreads. Now the Calutron Plant is used for stable 

radionuclides. In our TBD we were not able to get into how effective Y-12 was in controlling 

such contamination spreads, we just documented that this was a problem. 

SC&A: How effective have you been in characterizing off-batch processing operations and 

special campaigns like we brought up in our issue 13 in our questions list? It is not clear how 

NIOSH is handling dose reconstructions for these special campaigns. Medical isotope 

production is yet another area like this. 

ORAU: These kinds of issues are so specific and they are too small to have the opportunity to 

address. We try to address these for individual claimant’s but not so much in the TBDs. 

SC&A: NIOSH may want to rethink how you present these varying dynamics. The function 

ongoing in the building at each point of time is undergoing a lot of change. It is important that 

the TBDs provide the dose reconstructor with an idea of what is going on at specific times and in 

specific facilities. 

ORAU: As we get additional information, we try to do that. 

SC&A: In regard to question 13, actual operations at Y-12 could give rise to significant 

operational exposures while handling large amounts of 237Np and yet this is not addressed 

specifically in the TBD. 

ORAU: We did not find information yet documenting such 237Np handling operations at Y-12. 

In our discussion yesterday, during our external dose TBD conference call, we did address 240

Pu, 

and 239,240

Pu while discussing question 14. Under radionuclides of concern in Section 5.1.5 we 

did note that 237Np was a radionuclide of concern. The DRs have the latitude to look into the 

specifics. In the TBD we have only covered the primary ones. Some you can talk about and 

others you cannot. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

160 of 203 

SC&A: It seems that it will be important to learn more about these off-batch processing

operations since there is likely to be inadequate radiological controls in individual operations. 

NIOSH needs to better characterize these so as to better identify those which may have high 

exposures related to them. This is important for contaminated process streams as well. In our 

question 14, we also inquired about why plutonium likely to be of low “burnup” was not 

discussed. We note that Hanford operations involved production of 240

Pu. We wonder where it 

was going and if Y-12 did not receive a lot. 

ORAU: We presented this and discussed this at our Mel Chew meetings. It was discussed that 

240Pu would only be present in small quantities and thus would represent little dose impact. The 

TBD has taken a look at 238Pu and 239

Pu. 

SC&A: In regard to question 15, did not the handling of substantial quantities of tritium,

warrant a discussion in the Site Description TBD?

ORAU: As we mentioned during our external dose conference call yesterday, there was only a 

pilot program for tritium stated at Y-12 which ran for only a short time. That program was never 

implemented since they found that there were never significant quantities of tritium at the Y-12 

site and that doses to tritium were negligible. That is why the TBD is so sparse on this subject. 

SC&A: Did Y-12 ever store spent fuel? 

ORAU: There is no evidence that spent fuel was ever stored at Y-12. 

SC&A: In regard to question 16, in Section 2.4.6 it appears that new filters may not have 

operated effectively from their installation in 1955 through July 1956. What can you tell us 

about that? We note there have been several ventilation upgrades over the years and one is in 

process now. 

ORAU: Let’s defer that to our discussions on internal dose this afternoon. 

SC&A: Regarding question 17, it was noted that the first assessment of the S-3 ponds indicated 

that no special precautions were required for clean-up work. However, the second survey 

indicated that precautions were required. Was work on capping the ponds undertaken without 

such precautions between the two assessments? 

ORAU: We will have to look into that. 

SC&A: On page 20 of the TBD, two default solubility assumptions are give that comprise a 

mixture of Class M (90%) and Class S (10%). In interpreting the data, was consideration given 

to using a single solubility class intermediate between Classes M and M? On page 21, three 

default solubility assumptions are given for workers who participant exclusively in the urine 

bioassay program. Can you please comment on that?
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

161 of 203 

ORAU: Someone at Y-12 gave us this solubility information. We would have to ask him about 

this. IMBA can’t handle Class Q when you input your data. 

SC&A: Actually, IMBA can handle this by changing input data about the lung; you just have to 

know how to do it. A 50%S and 50%M might not work well for organ dose. It is hard to 

understand why Type S is used for lung dose and not Type M. 

ORAU: We will have to take this up with our Y-12 expert. 

SC&A: In our question 20, we noted that on page 21 of the TBD, Y-12 used a DOSE-66 code 

while IMBA is used in other technical basis documents SC&A has reviewed. DOSE-66 was 

developed in 1990 but was not validated until 2000. Why was the DOSE-66 given precedence? 

We need a reference on this? 

 

ORAU: DOSE-66 may be covered in the Y-12 internal dose TBD that Bryce Rich and Mel 

Chew published in May 2005. The original author, Howard Prichard, has passed away. 

SC&A: Who is the internal dosimetrist at Y-12 with who you consulted? 

ORAU: Mike Saulyiette is the ORNL dose we referred to. 

SC&A: In regard to our question 19, in Table 2.9-1 on page 21 of the Site Profile TBD, we note 

that all the uranium compounds are assigned solubility classes. Can you address these gaps?

ORAU: Several of these are not radioactive. 

SC&A: In regard to our question 21, in Table B.4-1, why are 238Pu and 239Pu assigned to 

different solubility classes? How is this reconciled with Table B.4-2?

ORAU: We would have to have our Y-12 internal dose expert this question. 

SC&A: In regard to our question 22, why is tritium vapor not discussed in the TBD text? 

ORAU: Because the Tritium Pilot Program was never finalized and they found dose potential 

from tritium to be negligible. 

SC&A: Table B.4-4 gives average data. Are the proportions of S and M available for each 

individual data point given so that a measure of variability can be obtained?

ORAU: This is left to judgment of the dose reconstructor. 

SC&A: Can you tell us more about the 86” Cyclotron?

ORAU: Look at the X-10 TBD for that information. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

162 of 203 

SC&A: Did Oak Ridge operate the Calutron at Y-12?

ORAU: See the discussion in the ORNL TBD. 

Y-12 Occupational Internal Dose TBD Conference Call Summary 

Date: June 8, 2005 

Subject: Occupational Internal Dose TBD, ORAUT-TKBS-0014-5 Rev. 01, 

 (Rich and Chew 2005) 

Time: 1:00 to 2:30 pm 

Individuals who participated included: 

ORAU: William Murray, Bryce Rich, Mel Chew, Elizabeth Brackett, and Ed Scalsky 

NIOSH: 

SC&A: Joe Fitzgerald and Joyce Lipsztein 

SC&A: The TBD (pg. 24) indicates that methodological problems occurred during the early 

period of uranium bioassay, resulting in an underestimate by a factor of two. How has 

NIOSH/ORAU accounted for this underestimation in the dose reconstructions? 

ORAU: All dose reconstructions to date have employed efficiency methods, so this issue would 

not be confronted until dose reconstructions require best estimates. Only 2 years, 1950–1951, 

are affected and the underestimate can be compensated through extrapolation back from 

succeeding years’ data. For earlier years, one can take the ratios of results recorded and use 

adjustment factors for correction factors. 

SC&A: The TBD (pg. 13) discusses process streams of uranium. Included in the process 

streams are something identified as “solvent extraction raffinates.” What are “solvent extraction

raffinates”? Is this a process that may concentrate particular radionuclides including daughter 

products?

ORAU: These are metallic impurities removed by chemical columns designed to remove metals 

from uranium solutions. Certain impurities, such as trace radionuclides, may be concentrated by 

this process and this is captured through the default ratios used in calculations. 

SC&A: The chemical forms of uranium at Y-12 include S, M and F type materials. What is the 

basis for the use of Type M material from 1948-June 1998 without regard to the type of cancer?

ORAU: This assignment was based on information from site records. Uranium at Y-12 is likely 

to be either Type S or M. A variety of dose modeling curves are available based on solubility, 

with ORAU using the most claimant favorable. 

SC&A: Provide an example of how the recycled uranium impurities are accounted for in the 

dose reconstruction. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

163 of 203 

ORAU: Rev 01 of the TBD does include recycled uranium impurities for purposes of dose 

reconstruction. Two defaults are commonly used: best estimates for long-term exposure; and 

maximum credible at the 95 percentile CL. 

SC&A: In the TBD (pg. 23) there is a reference to Section 5.4.2 for discussion of missed doses. 

This section appears to be missing. Is it going to be included in the near future? The TBD 

mentions that there is a concern that the technique used before May 1952 could have 

underestimated the urinary uranium concentrations. How is the dose reconstructor going to deal 

with these underestimates of urinary uranium concentrations? 

ORAU: Yes, both sections 5.4.1 and 5.4.2 were deleted in this revision of the TBD, but 

references to these sections were inadvertently left in. More detail for the dose reconstructor will 

be included in the next revision which is underway. 

SC&A: Was tritium dose included in the whole-body dose at Y-12?

ORAU: Yes, but it was not present in any appreciable quantities at Y-12. 

SC&A: Is NIOSH considering the issue of frequency of monitoring during the remediation of 

the S-3 ponds (Section 2.6 of the Site Description TBD, ORAUT-TKBS-0014-2 (Jessen 2005)? 

Has NIOSH investigated the incidents that occurred or the potential for acute exposures that may 

have existed during the remediation of the S-3 ponds?

ORAU: ORAU acknowledged that two different evaluation reports for the S-3 pond 

remediation were referenced which apparently contradict each other regarding the significance of 

potential radiation exposures. 

SC&A: How does NIOSH plan to address early period internal dose issues?

ORAU: There is no discernable Eastman data for the 1943–1946 time period. OTIB 29 – relies 

on internal dose co-worker data for 1947–1988 time period; there is a need to extrapolate back 

results because of underestimates back to Calutron shutdown. No internal dose data exists 

before 1950. 

SC&A: What were the processes involving 232

Th at Y-12? 

ORAU: ORAU has reviewed Y-12 ledgers for thorium receipts. It came in as a “gel,” pressed 

into forms by machinery, and rolled in the DU facility. Y-12 had been sensitive to the need for 

precautions in handling thorium; monitoring included routine lung counting and air monitoring. 

It was processed extensively during 1962–1976, with limited processing from 1985–1988. Work 

done in the early 1950s was limited to research and development. 

SC&A: On page 7 of the TBD, it mentions that no urinalysis data have been found for Y-12 

prior to 1948. Is there any explanation for why no urinalysis data have been found prior to 1948? 

The fluorometric technique should have been well established by that date. What steps are in 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

164 of 203 

 

hand to resolve this issue? Does the discussion of lack of urinalysis prior to 1948 (p. 5, vol. 5) 

mean that NIOSH is not doing dose reconstruction for the early period at this time? 

ORAU: Eastman statements were found in old papers that acknowledged the use of

fluorometric procedure for uranium dosimetry. However, no dose records have been located. 

SC&A: Page 8 of the TBD: From the earliest days, it was recognized that ‘insoluble’ uranium

(except for high-fired oxides) exhibited a solubility intermediate between Classes W and Y. This 

was addressed in one way by defining Class Q. However, was any attempt ever made to develop 

a model specifically for these uranium compounds and to then interpret the Y-12 data explicitly 

with that model? More generally, this report addresses joint interpretation of the urinary and 

fecal data, but says nothing about interpreting these data in conjunction with the chest counting 

data. Has such an overall analysis been attempted either at the level of the individual or in 

relation to groups of workers with similar exposure histories?

ORAU: The Q class was used briefly at Y-12, but is no longer used. IMBA can handle class Q, 

if need be. For the 1950s, co-worker data is relied upon. 

SC&A: Page 9 of the TBD: Figure 5-2 gives a composite urinary excretion curve for 157 

individuals. This conforms quite well to Type M behavior. However, this is not unexpected at a 

plant where a wide variety of uranium compounds of different degrees of solubility were used. It 

would be much more interesting to see the curves for the individual workers to see if they span 

the full range from Type F to Type S (or even beyond). Is this possible?

ORAU: No. NIOSH does not think it is useful, because workers worked at several different 

places, with several different jobs, thus being exposed to all solubility types 

SC&A: Page 10 of the TBD: The particle sizes range up to large values (10 μm physical 

diameter). Are there particle size data for respiratory zone relative to general area monitoring 

and can any useful relationships be established between the two?

ORAU: The workers were assigned to different work locations and thus we do not think it is 

useful to try to characterize the particle sizes of different processes and areas. NIOSH uses the 

default 5 μm for particle size.1

 

SC&A: Page 11 of the TBD: There is some indication in the TBD’s Appendix that distinctions 

were made between routine and special monitoring. It seems likely that special monitoring 

would have been triggered by individual incidents. Could an account be given of the special 

monitoring program employed, including triggers for special monitoring, duration of follow up 

and any analyses of the results that have been undertaken?

ORAU: We do not think this is useful data. 

1

 Although not discussed during the conference call, there are other particle size distributions that could 

result in higher doses. This issue is addressed in the main body of this review. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

165 of 203 

SC&A: Page 20 of the TBD: Can the constant “8” in equation 3.1 be reconstructed based on a 

detailed knowledge of the assay procedure used? 

ORAU: No, there is no information. 

SC&A: Page 21 of the TBD: Here and elsewhere, a nominal counting efficiency of 0.5 

cpm/dpm is used. Bearing in mind that this is for counting of a disk, is this the maximum

possible value that can be achieved.? This seems implausible. Can details of the counting 

geometry be provided, so that this issue can be examined?

ORAU: We agree that this counting efficiency is not plausible. 

SC&A: Page 22 of the TBD: For the recent data on uranium in urine by alpha spectrometry, 

have comparisons been made with other methods, e.g. inductively coupled mass spectrometry 

(ICPMS) or accelerator mass spectrometry (AMS)? More generally, how well are the various 

assay procedures used verified by inter-laboratory comparison exercises?

ORAU: It is not our job to verify inter-comparison of techniques. 

SC&A: Page 27 of the TBD: Fecal sampling is effective at recording recent intakes of

insoluble materials by ingestion and inhalation. Has consideration been given to whether fecal 

samples obtained at an interval of 53 days give a meaningful representation of chronic intakes 

since there is considerable temporal variability? 

ORAU: We cannot explain the 53 days interval for obtaining fecal samples. We know that this 

frequency comes from an analysis done by (Eckerman and Kerr 1999), and it is explained in the 

Y-12 bioassay documents, that have been retrieved by SC&A. 

SC&A: Page 27 of the TBD (Interferences): The methodology used to adjust results for 

background is different from the one recommended by (Eckerman and Kerr 1999) for the Y-12. 

Why?

ORAU: NIOSH will not adjust for background. We will probably take out this section on the 

next revision of the document. 

SC&A: Page 35 of the TBD: From what source can the referenced document (Wilcox 1999) be 

obtained? 

ORAU: We will provide that document to SC&A. 

Internal Dosimetry Co-worker Data 

SC&A: This report appears to, first, average urinary excretion data over individuals, and then 

interpret the results in terms of Type M and Type S intakes. Can NIOSH clarify why this was 

done, as the results must reflect different numbers of workers in different contexts? Also the 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

166 of 203 

results will inevitably be skewed in time because of the different time courses of exposure of the 

individual workers, as each worker will have been associated with multiple determinations of

uranium in urine. Why was this analysis not undertaken on an individual-by-individual basis? 

This would have yielded a much richer understanding of the range of exposure conditions. 

ORAU: It is impossible to do this analysis on an individual–by –individual basis, because of the 

enormous amount of data. 

SC&A: Given that multiple intake periods that were used, the fitting model should be able to 

give a very accurate representation of the data. However, there are significant divergences for 

the Type M result (Figures A-9 and A-10) of the TBD, and these are much worse for the Type S 

analysis (Figures A-19 and A-20). These divergences arise from fitting data for each of the 

intervals separately and then adding the results. Why was this done instead of undertaking a 

simultaneous joint fit to all the intervals, or at least rescaling each of the individual fits by an 

overall normalization factor to give a better overall fit? (However, this remark should be seen in 

the context of the previous consideration that an individual-based analysis is to be preferred to 

the group analysis presented here). 

ORAU: As we said above, it is impossible to do this analysis on an individual–by –individual 

basis, because of the enormous amount of data. 

SC&A: OCAS-OTIB-0029, Internal Dosimetry Coworker Data for Y-12 (Brackett 2005) 

calculates the 50th and 84th percentiles for urinalysis data. What is the basis for choosing the 84th

percentile? 

ORAU: In order to obtain the standard deviation. 

SC&A: Many of the tables and graphs in OCAS-TIB-0029, Internal Dosimetry Coworker Data 

for Y-12 are missing the time period from May 1, 1952 to July 31, 1953. Is this intentional, or 

should this time period appear in all relevant tables and graphs?

ORAU: They should not be missing. 

SC&A: Did NIOSH take into consideration that urine samples were collected on a Monday 

morning, after a minimum of a 48-hour absence from the work area?

ORAU: We did not take into consideration the 48 hours absence from work area. We will verify 

if results would have been different if we had assumed a 48 hours absence from work area. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

167 of 203 

ATTACHMENT 5: SUMMARY OF ENVIRONMENTAL 

CONFERENCE CALL WITH NIOSH AND SC&A 

Date: June 10, 2005 

Subject: Occupational Environmental Dose TBD, ORAUT-TKBS-0014-4, Rev. 00, PC-2 

(Ijaz and Adler 2004) 

Time: 1:30 - 2:30 pm

Individuals who participated included: 

ORAU: Ed Scalsky, Tim Adler, Mel Chew, Bryce Rich, Bill Murray, Talaat Ijaz 

NIOSH: Stu Hinnefeld 

SC&A: Bob Anigstein 

This summary first presents the original question or comment, identified by number, that was 

distributed in advance to ORAU and NIOSH, as it appeared in the original transmittal. This is 

followed by a summary of the discussion of each question. The speakers are identified by the 

organization they represent. The cited statements are condensed summaries or paraphrases 

rather than verbatim quotes. 

1. The TBD addresses only two exposure pathways: inhalation of uranium in ambient air 

and external exposure to direct penetrating radiation. Whenever individuals perform

outdoor work, there is a potential for ingestion of soil and other windblown particulates 

that are larger than the respirable particles. This pathway is traditionally included in dose 

assessments (see EPA Exposure Factors Handbook@). What is the basis of neglecting 

inadvertent ingestion of radioactively contaminated soil and other finely dispersed 

radioactive materials?

SC&A: Was there a scoping evaluation? Was it considered?

ORAU: TBD was based on CDC dose reconstruction, which did not address 

inadvertent ingestion. TBD parameters only list inhalation & external 

exposure. Someone higher up decided that dust resuspension & inadvertent 

ingestion were not pathways to be addressed. 

NIOSH: I think it would be a relatively minor contribution to exposure. It may be 

possible to evaluate magnitude. If I'm wrong, we would have to address that. 

ORAU: Resuspension also not considered because Y-12 site consists of buildings, 

asphalt parking lots, and roads. Resuspendable material would have been 

washed away. Vegetative cover near the southern boundary, not a significant 

area. 

SC&A: What about off-site deposition that could be blown on site?
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

168 of 203 

ORAU: Mostly highly wooded areas that would have required significant winds to 

generate clouds of dust that would have been blown on site. Quite minor 

pathway for uranium. 

2. There are several references in this TBD to monitoring of fission products. What is the 

basis of neglecting inhalation of radionuclides other than uranium?

SC&A: Since there was monitoring for other nuclides, including fission products, 

shouldn't they have been addressed in the TBD?

ORAU: We relied heavily on the CDC dose reconstruction, which did screening 

assessment for Y-12. Screening assessment, done for off-site dose, included 

Pu and Np, but concluded U only nuclides of concern. 

NIOSH: We will pursue it further with ORAU. 

3. The TBD states that uranium air concentrations were not available after 1999. However, 

since TDEC continued to collect samples and analyze them for gross α and gross β, 

uranium air concentrations could be estimated from these measurements, using the ratios 

of uranium concentrations to the gross α and gross β measurements made in earlier years. 

What is the basis of excluding these data from developing the guidance for dose 

reconstruction?

SC&A: Was it not possible to prorate TDEC gross alpha from previous years to give 

U activities for later years?

ORAU: TDEC measurements were made in μg/m3

. Using these data would have 

required knowledge of isotopic ratios. If total U exceeded criteria, TDEC 

would then do isotopic analysis. 

SC&A: TBD states that measurements were gross α and gross β. Same isotopic mix 

could be assumed so as not to ignore data. 

Back & forth discussion on TDEC measurements. 

 

SC&A: Observation: TDEC measurements should be made available to dose 

reconstructors. 

Back & forth discussion on methods to be used to reconstruct doses for years after 1995. 

4. Data regarding the AMAD or the chemical form of the airborne uranium is required by 

dose reconstructors, who need to assign dose coefficients to the uranium isotopes. Has

NIOSH attempted to determine the AMAD and chemical form of these particulates? 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

169 of 203 

 

SC&A: Chemical form and AMAD should be furnished. Indoor default is 5 μ,1

 but

outdoor is often 1 μ. 

ORAU: The original scope of work was to provide air concentrations. 

NIOSH: Presume there is no or insufficient data on chemical form or solubility class. 

DR's2

 use the most favorable solubility class, probably S or M, for a specific 

claim. Uncertain about particle size. Typically use 5 μ, since pollutants 

generated in industrial setting. 

SC&A: Why not compare 1 μ and 5 μ AMAD and use the more claimant-favorable 

assumption? 

NIOSH: We haven't typically done that. 

ORAU: Particle size was not measured. 

5. The empirical χ/Q approach is based on several premises which are not explicitly stated 

and which need to be explored and justified before this approach can be accepted. 

(a) All the occupationally-related airborne activity originates at Y-12. However, Y-12 is 

part of the ORR complex, which includes other sources of radioactive material. The 

varying correlation of the airborne uranium concentrations to releases from Y-12, 

especially at Station 12, calls this hypothesis into question. Has NIOSH considered 

any alternate approaches to estimating the airborne uranium concentrations? 

(b) In Section 4.2.5, the TBD attributes the poor correlation at Station 12 to the paucity 

of data. The smaller number of data points (8 vs a maximum of 17) does not account 

for the virtual lack of correlation shown in Figure 4.2.5-4. We believe that fewer 

data points would lead to a better correlation, both on mathematical grounds (fewer 

degrees of freedom), and on physical grounds: meteorological conditions would 

vary less during a shorted time period. Has NIOSH considered eliminating the 

Station 12 data in creating a framework for dose reconstruction, due to this lack of 

correlation? 

(c) Resuspension of uranium deposited on soil, both inside the Y-12 perimeter and off 

site, is not included. Resuspension of deposition during previous years would 

contribute to the uranium concentration in air, yet is uncorrelated to uranium releases 

from Y-12. Has NIOSH considered adjusting the airborne uranium concentrations 

by estimating the activity in the soil, perhaps by employing the surveys of external 

exposure rates, and employing a resuspension factor?

1

 Micron (μm)

2

 Abbreviation for Adose reconstructors.@
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

170 of 203 

(d) The influence of other sources of uraniumCreleases from sources outside Y-12 or 

resuspension from on-site soilsCis indicated by the data for 1993. The uranium

release of 3 kg for that year is the lowest during the period that Stations 2 and 12 

were operating, and the second lowest during the 17-year operating period of 

Stations 4 and 8. The 1993 calculated χ/Q's for all four stations are the highest of 

any year that these stations operated. At least for the data from Station 12, which 

SC&A subjected to an independent statistical analysis, there is a negative correlation 

between the χ/Q's calculated by ORAU and the annual releases of uranium: the 

lower the releases, the higher the χ/Q. How does NIOSH plan to resolve this issue?

SC&A: We are somewhat skeptical about the empirical χ/Q approach. Has any other 

approach been considered?

ORAU: This goes back to the original scoping assessment. Atmospheric modeling 

was done for all 3 sites on the ORR. Contribution from other 2 facilities to Y12 is minimal at best. There are problems with atmospheric modeling at Y-12 

due to the complex terrain. 

SC&A: Agree that analytical approach is difficult and questionable. However, 

correlation between releases and concentrations is poor in some cases. Note 

that the year there is the smallest releaseC3 kgCfrom Y-12, there is also 

consistently the highest χ/Q. Presumably, if you had zero releases, the 

concentrations would be far above zero. This indicates that there were 

contributions from other sources. 

ORAU: Except for Station 12, correlations are from 0.95 B 0.7. Don't agree that 

correlations are that poor. 

 

SC&A: Overall scatter is fair. Station 12 with correlation 0.19, should not be used. 

Not scientifically valid. For every station, 1994, with the smallest release, 

produced the highest χ/Q. Unless you believe in a coincidence of wind 

patterns that cause higher concentrations at all 4 stations, this is a clear 

indication that there were other sources that are being neglected. 

ORAU: That's for just one year, right? 

SC&A: That one year emphasizes it, because the release is so small. If you go to the 

highest year, so if 1984 has 300 kg vs 3 kg, yet the concentration does not 

track that. Anyway, that's an observationCthis may be an estimate, but we 

question if it's really claimant favorable 

. 

ORAU: Is that observation based on that one year?
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

171 of 203 

 

SC&A: That one year brings the whole concept into question. The concept is that 

releases from the Y-12 plant are correlated with the concentrations. There 

should be some adjustment factors to make it more claimant favorable. 

6. The TBD states that the calculated χ/Q's follow a lognormal distribution with a high-end 

tail. Yet, in assigning 95th percentile values, where the high-end tail would have the 

greatest influence, that aberration from lognormal behavior is not addressed. An 

examination of the data in Tables 4.2.3-1/4 reveals the following: 

(a) Station 2: out of 12 calculated χ/Q's, one is significantly higher than the 95th

percentile value (2.00 H 10-14 vs. 1.31 H 10-14). Statistically, the highest of 12 values 

would correspond to the 92nd percentile. 

(b) Station 4: out of 17 calculated χ/Q's, one is over 1.7 times higher than the 95th

percentile value (5.03 H 10-14

 vs. 8.73 H 10-14). Statistically, the highest of 17 values 

would correspond to the 94th percentile. 

(c) Station 8: out of 17 calculated χ/Q's, two are higher than the 95th percentile value. 

Statistically, the second highest of 17 values would correspond to the 88th percentile.

(d) Station 12: out of only eight calculated χ/Q's, one is higher than the 95th percentile 

value. Statistically, the highest of eight values would correspond to the 88th

percentile. 

The above observations indicate that the distributions are not lognormal, and that the 

assumptions used by ORAU to calculate the 95th percentile χ/Q's are not claimant 

favorable, aside from the questions raised earlier in this discussion about the validity of 

the empirical χ/Q approach. This conclusion is further buttressed by our statistical 

analysis of the Station 12 data, which exhibit the best correlation between the annual 

releases and the annual average air concentrations of uranium of the four monitoring 

stations. Subjecting the calculated χ/Q's to the W test3

 shows that there is a 10% to 50% 

probability that the distribution is lognormal. In other words, there is a better than a 50% 

chance that the distribution is not lognormal. Has NIOSH considered an alternate 

statistical approach that is more claimant favorable? 

SC&A: Not sure if DR is supposed to use 95th percentile. 95th percentile calculated 

from lognormal distribution is not, in fact, the 95th percentile of the data 

because of the high-end tail. 

 

ORAU: Our conclusion was the same as yours. The correlation to a lognormal 

distribution was very vague and questionable at best and was not included in 

the initial report. However, DRs had a requirement that the data fit some sort 

of distribution. In the end, the decision was made by DRs to provide 95th

3

 Richard Gilbert (1987). Statistical Methods for Environmental Pollution Monitoring, Section 12.3.1.
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

172 of 203 

percentile, knowing full well that the distribution of the data did not fit well 

with a lognormal distribution. 

 

SC&A: It is our understanding that IREP does not require a distribution. It will accept 

single values. A single value could have been assignedCsimply the highest. 

Given that you have 17 data points, the highest of those could have been 

represented as the 95th percentile. 

ORAU: If you look at Table 4.2.3-1 through -4, the maximum value is printed. 

 

SC&A: I know it is. 

ORAU: The selection of the value to be used was not our recommendation, it was the 

recommendation of the DRs. 

 

SC&A: Can NIOSH address that?

NIOSH: From a DR standpoint, you can choose a high value that would be an upper 

end of what could occur. The difference between selecting the 95th percentile 

and the highest value, dosimetrically is not going to be consequential. 

 

SC&A: It can make a 2- to 3-fold difference. We realize that the DR has latitude, but 

the idea of the TBD is to provide them with direction. Our recommendation 

would be that the direction be more clear, rather than leaving it to the 

individual DR. 

7. NIOSH decided to base the dose reconstruction on the calculated χ/Q's, even for years for 

which measurements of air concentrations exist. Since the purpose of dose 

reconstruction is to determine the doses to individual claimants, not to maintain a 

uniformity of approach, what is the justification for disregarding real data, especially in 

cases where the use of such data may be more claimant favorable?

SC&A: If you have years where you have air measurements, that would be the more 

reliable data. The idea is to have the best dose for each claimant, rather than 

have artificial consistency. Our comment would be to that NIOSH should 

reconsider if the DRs should not be directed to use the actual air 

concentration. To disregard measured data in favor of a fitted curve doesn't

seem to be claimant favorable or scientifically valid. 

NIOSH: Suppose you had a year where the measured data was lower than the 

calculated χ/Q value?

SC&A: I would use whichever one is the more claimant favorable between the two, or 

use the measured data and say the calculated values should only be used in the 

absence of measured data. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

173 of 203 

NIOSH: Either way, you feel that would be an acceptable approach?

SC&A: Yes, you can't argue with the fact that it's measured data, but you should put 

an uncertainty on it. 

8. The methodology for calculating the air concentrations of the uranium isotopes, which 

are listed in Attachment D to the TBD, is unclear. Our attempts to replicate some of the 

data in Attachment D, using the information in the main body of the TBD, were 

unsuccessful. Furthermore, some of the data in Table 4.2.4-1, which are the basis of the 

data in Attachment D, are themselves suspect. Notably, based on the masses of the 238

U 

releases listed in the fifth column of that table, the 238U activities released in 1994 and 

1995 should be 0.008 and 6.7 H 10-4 Ci, respectively, rather than 0.002 and 0.0021 Ci, as 

listed. Would NIOSH clarify the methodology used in constructing the Attachment D 

tables, and verify the data in Table 4.2.4-1?

SC&A: I could not reproduce the numbers in tables in Attachment D. Perhaps we 

could get an example calculation. 

 

ORAU: Yes. 

9. On p. 26, the TBD states: A. . . it is conservative to assume that the air concentrations

reported from 1996 to 2002 are equal to the concentrations reported for 1995.@ However, 

there are no air concentration data listed for the years 2000 B 2002. The TBD states that 

the data for these years, collected by TDEC, is not usable. More important, the statement 

quoted above is contradicted by the data that are listed. The annual average 

concentrations at Station 4 for the years 1996 B 1998 are higher than for 1995. 

Furthermore, the data in Table 4.2.4-1 shows that the uranium releases in each year from

1996 through 2001 were higher than in 1995. Can NIOSH justify the assumption quoted 

above?

SC&A: There are data for Station 4 for 1996 - 1999. 

ORAU: The enrichment is needed to calculate isotopic ratios. 

 

SC&A: The releases in 1995 are the lowest for the years 1944 - 1995, a total of 2 kg 

U. However, 1996 - 2001 are all higher than 1996. 

ORAU: Figure 4.3-2 shows decreasing releases. 

SC&A: The data in the figure end with 1995. We recommend increasing the release 

estimates for 1996 - 2001 in proportion to the total releases in those years, or 

performing some other adjustment. 

10. The tables in Attachment D list both 50th and 95th percentile values. Presumably, the 

50th percentiles are to be used as the best estimates, and the 95th percentiles as the upper 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

174 of 203 

bound values. As mentioned earlier, the calculated 95th percentile values are not upper 

bounds, as they are exceeded by annual average χ/Q's at each of the four monitoring 

stations. Can NIOSH explain how these data are to be used by dose reconstructors?

[This question had been discussed earlier.] 

11. Concentrations of 236U, which are listed in Tables 4.2.3-1/4, are comparable to those of 

238U. Why are the contributions of 236

U ignored in Attachment D? 

 

ORAU: We used the same isotopes as the CDC dose reconstruction, which did not 

include 236U. 

SC&A: Agree that 236

U would not make a large contribution to the dose, but from a 

scientific standpoint, it should be addressed. 

 

12. The TBD states that the aerial surveys performed by the EG&G Remote Sensing 

Laboratory cannot be used for dose reconstruction. However, by spanning a period from

1973 to 1989, they provide a historical record of the external exposures on the site. Such 

data could be used to provide a temporal trend, and should be combined with the 1985 B

1987 scoping survey. The spatial resolution of the aerial surveys is comparable to the 

size of the grid blocks in the scoping survey. Furthermore, the exposure rates calculated 

at 1 m above ground, as presented in the aerial survey reports, appear to be more relevant 

to dose reconstruction than the ground level rates measured in the scoping survey. Has 

NIOSH compared the dose rates based on the aerial surveys to those of the scoping 

survey, and has it considered developing a time-dependent dose rate based on both sets of 

data?

ORAU: There were problems with the legibility of the black & white copies of the 

aerial survey data. We heard comments from Oak Ridge staff about 

discrepancies between data collected by a helicopter and a fixed-wing aircraft. 

SC&A: The original full color maps can be purchased from the RSLCwe had, in fact, 

obtained these maps for an earlier study. We recommend that the aerial 

survey data be used to supplement the ground survey. 

13. Given the purpose of the survey and the instruments used, the exposure rates would have 

been based on the count rates registered by the sodium iodide crystal in the survey meter. 

Such meters are commonly calibrated with a 137

Cs sourceCthe scale on the rate meter is 

set to the known exposure rate at the location of the detector. This calibration is thus 

valid only for the principal 137

Cs γ ray and will vary significantly for photons of different 

energies. Has NIOSH attempted to convert the meter readings to dose rates in air (i.e., air 

kerma), using calibration curves which can be obtained from the manufacturer of the 

meter in question, or from manufacturers of similar instruments, and the known spectrum

of the uranium isotopes deposited on the ground? Has it considered using the spectral 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

175 of 203 

analysis in the aerial survey data, and then converting the air kerma to dose equivalent 

rates, using tables in ICRP Publication 74 for the appropriate exposure geometry?

ORAU: The ground survey data was all that was available. 

 

SC&A: You could use the energy-dependent calibration curves to convert the 

readings, based on the known energies of the isotopes in question. 

 

NIOSH: Did the survey over- or under-estimated the actual concentrations? 

SC&A: The uranium concentrations would most likely be over-estimated, while 

nuclides with gamma energies > 660 keV would be under-estimated. 

The conference ended with ORAU agreeing to e-mail the example calculations for Attachment D 

to SC&A early next week.
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

176 of 203 

ATTACHMENT 6: SITE EXPERT INTERVIEW SUMMARY 

<RESERVED> 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

177 of 203 

ATTACHMENT 7: CONSISTENCY BETWEEN Y-12, HANFORD AND 

SRS SITE PROFILES 

Table A.7.1 Occupational Medical Exposure Default Assumption Comparison for Y-12, 

Hanford and the Savannah River Sites 

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS

Frequency of chest x-rays

(Default) 

1943 – Present Pre-placement & 

At termination

1943 - July 18, 1988, annually 

July 18, 1988 – present 

Time of entry into asbestos and 

beryllium areas 

July 18 – March 1998

Active & Previous beryllium

workers 

July 18, 1988 – June 30, 1993

Every 10 yrs – Under 30

Every 5 yrs – 30–45 y/o

Every 3 yrs – Over 45

March 1998 – present

Annually – all asbestos workers 

Every 3 yrs – previous and 

active asbestos workers and 

previous beryllium workers 

(Murray 2004, pg. 15) 

Posterior-Anterior View: 

Before 1946 – 1/1982: 

Pre-employment, annual, 

and termination 

1/82-1/83: Pre-employment, annual, and

termination for over 

50 years; Biennially for 

40-49 years; Every third 

year for 39 years or

younger.

1/83–3/90: Biennially for 

over 50 years; Every third 

year for 40–49; and Every 

five years for 39 years

and younger. 

3/90–present: Every 

5 years 

Lateral chest x-rays also

given periodically prior to

4/1997. 

One annual x-ray 

procedure for each 

year or partial year. 

Organ Dose Conversion

Factors 

Obtained from ICRP 34 (1982)

for organs outside the chest

cavity (Murray 2004, pg. 15)

Obtained from ICRP 34 

(1982)

Obtained from ICRP 

34 (1982) 

Substitute dose conversion

factors for thyroid,

eye/brain, ovaries and 

analogues, testes, and 

uterus 

Doses for organs not listed in

ICRP 34 but specified in IREP 

code are determined by analogy

and anatomical location as 

indicated in Table 3.5, pg. 9 

(Murray 2004) 

Use DCFs for lung for all 

other organs in thoracic 

cavity; for organs in

abdomen, use DCFs for

the ovary (Scalsky 2003, 

pg. 10) 

Use substitute dose 

conversion values 

outlined in Table 5.1-1,

ORAUT-OTIB-0006 

Use substitute dose 

conversion values 

outlined in Table 5.1-1,

ORAUT-OTIB-0006 

(also Table 2.6, SRS 

TBD) prior to 1970

(Scalsky 2004, pg. 50). 

IREP Radiation Rate Acute Acute Acute 

IREP Radiation Type Photons, 30-250 keV Photons, 30-250 keV Photons, 30-250 keV

IREP Dose Distribution 

Type 

Constant Constant Constant

Total uncertainty 30%, pg. 10 (x-ray dose 

multiplied by 1.3 and entered as 

a constant 

30% (x-ray dose 

multiplied by 1.3 and 

entered as a constant) 

30% (x-ray dose 

multiplied by 1.3 and 

entered as a constant) 

Conversion Factor from PA

to Lateral 

Kathren et al, to be published,

pg. 7 

2.5 2.5
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

178 of 203 

Table A.7.1 Occupational Medical Exposure Default Assumption Comparison for Y-12, 

Hanford and the Savannah River Sites 

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS

Chest Thickness PA indicates a posterior/anterior 

view, the average PA chest 

measures 26 cm, pg. 7 

Standard chest thickness 

is 22-24 cm. For chest 

thicknesses of 25-27 cm, 

increase the dose by a

factor of 1.5. For chest

thickness of >27 cm, 

increase the dose by a

factor of 2. (Scalsky 

2003, pp. 8 & 17)

PA View: 26 cm

Lateral View: 34 cm

No corrections for 

varying chest

thicknesses are

mentioned in the TBD.

Analogue organ for 

Thymus 

Lung, pg. 9 Lung Lung 

Analogue organ for 

Esophagus 

Lung Lung Lung

Analogue organ for 

Stomach 

Lung Lung Lung

Analogue organ for Bone

Surface 

Lung Lung Lung

Analogue organ for Liver,

gall bladder, spleen 

Lung Ovary Lung

Analogue organ for 

Remainder Organs

Lung Ovary Lung

Analogue organ for 

Urinary/bladder and 

colon/rectum

Ovary Ovary Ovary

Analogue organ for 

Eye/brain 

Thyroid Thyroid Thyroid

Skin dose Skin dose is entrance skin 

kerma, multiplied by a 

backscatter factor of 1.35 from

NCRP 102, Table B-8, pg. 17

and 18 footnote (Murray 2004)

Skin dose was determined

by multiplying the ESE 

by the backscatter factors 

of 1.35 and 1.4 for HVLs 

of 2.5 and 3.5 mm Al, 

respectively (Scalsky 

2003, pg. 158). 

Skin dose was

determined by

multiplying the ESE by

the backscatter factors 

of 1.35 and 1.4 for 

HVLs of 2.5 and 3.5

mm Al, respectively 

(Scalsky 2004, pg. 

158). 

Posterior-Anterior View X-ray Techniques1,2, 3

<1946 1943–1947 kVp: 80, pg. 7

mAs: 200 

SSD: 2.95” (7.5 cm), pg. 10

kVp: Unknown

mAs: Unknown 

SSD: 72” (183 cm) 

ESE: 120 mR

Not Applicable 

2/1946 – 12/ 1950 1948- 1968 kVp: 80

mAs: 200 

SSD: 2.95” (7.5 cm) 

ESE: Not specified

kVp: 80

mAs: 25 

SSD: 72” (183 cm)

ESE: 79 mR

Not Applicable unless 

pre-employment is

indicted. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

179 of 203 

Table A.7.1 Occupational Medical Exposure Default Assumption Comparison for Y-12, 

Hanford and the Savannah River Sites 

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS

1/19/51 - 1/1/57 2/1948 – present kVp: 80 

mAs: 200 

SSD: 2.95” (7.5 cm), pg. 10

ESE: Not specified

kVp: 80

mAs: 10 

SSD: 72” (183 cm)

ESE: 79 mR

kVp: 80

mAs: 30 

SSD: 152 cm

No Collimation

ESE: 108 mR

1/1/57 – 4/1959 2/1948 – present kVp: 80

mAs: 200 

SSD: 2.95” (7.5 cm), pg. 10

ESE: Not specified

kVp: 80

mAs: 10 

SSD: 72” (183 cm)

ESE: 79 mR

kVp: 80

mAs: 30 

SSD: 152 cm

No Collimation

ESE: 108 mR

4/1959 – 12/1970 2/1948 – present kVp: 80

mAs: 200 

SSD: 2.95” (7.5 cm), pg. 10

ESE: Not specified

kVp: 80

mAs: 10 

SSD: 72” (183 cm)

ESE: 40 mR

kVp: 80

mAs: 30 

SSD: 152 cm

No Collimation (prior 

to 1970)

ESE: 108 mR

1/1971 – 1/1983 2/1948 – present kVp: 80

mAs: 200 

SSD: 2.95” (7.5 cm), pg. 10

ESE: Not specified

kVp: 80

mAs: 10 

SSD: 72” (183 cm)

ESE: 40 mR

kVp: 110-120 

mAs: 10 

SSD: 152 cm

ESE: 44 mR

1/1983 – 7/1985 2/1948 – present kVp: 80

mAs: 200 

SSD: 2.95” (7.5 cm), pg. 10

ESE: Not specified

kVp: 100 

mAs: 10 

SSD: 72” (183 cm)

ESE: 35 mR

kVp: 110-120 

mAs: 10 

SSD: 152 cm 

ESE: 44 mR

8/1985 – 3/1990 2/1948 – present kVp: 80

mAs: 200 

SSD :2.95” (7.5 cm), pg. 10 

ESE: Not specified

kVp: 100 

mAs: 10 

SSD: 72” (183 cm)

ESE: 35 mR

kVp: 120 

mAs: 7.5 

SSD: 152 cm

ESE: 33 mR

Posterior-Anterior View X-ray Techniques1,2. 3

3/1990 – 4/1997 2/1948 – present kVp: 80 

mAs: 200 

SSD: 2.95” (7.5 cm), pg. 10

ESE: Not specified

kVp: 110 

mAs: 6.7 

SSD: 72 “ (183 cm)

ESE: 21 mR

kVp: 120 

mAs: 7.5 

SSD: 152 cm

ESE: 33 mR

4/1997 – 2/1998 2/1948 – present kVp: 80

mAs: 200 

SSD: 2.95” (7.5 cm), pg. 10

ESE: Not specified

kVp: 110 

mAs: 10 

SSD: 183 cm

ESE: 17 mR

kVp: 120 

mAs: 7.5 

SSD: 152 cm

ESE: 33 mR

2/1998 – 5/1999 2/1948 – present kVp: 80

mAs: 200 

SSD: 2.95: (7.5 cm), pg. 10 

ESE: Not specified

kVp: 110 

mAs: 5

SSD: 183 cm

ESE: 11 mR

kVp: 120 

mAs: 7.5 

SSD: 152 cm

ESE: 33 mR

5/1999 – present 2/1948 – present kVp: 80

mAs: 200 

SSD: 2.95” (7.5 cm), pg. 10

ESE: Not specified

kVp: 110 

mAs: 5

SSD: 183 cm

ESE: 11 mR

kVp: 120 

mAs: 7.5 

SSD: 152 cm

ESE: 33 mR
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

180 of 203 

Table A.7.1 Occupational Medical Exposure Default Assumption Comparison for Y-12, 

Hanford and the Savannah River Sites 

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS

Photofluorography

Technique Factors X-rays were done at Oak Ridge 

Hospital or Y-12, Techniques 

are unknown. pg. 7 

kVp: 80 to 100 kVp 

mAs: not specified 

SID: 102 cm

ESE: 1.53 R 

Applies 1945 to 1962

kVp: 100 

mAs: 60 

SID: 102 cm

ESE: 1.5 R 

Applies from 1951–

1957

1 Refer to Scalsky 2004, pp. 41-47 for SRS x-ray technique discussion. 2 Refer to Murray 2004, pg. 18 for Hanford x-ray technique summary. 3 Available data indicate the x-ray beams used at Hanford were collimated. (Scalsky 2003, pg. 8). 4 N/A = not applicable; PA = posterior-anterior; LAT = lateral; kVp = kilovolt potential; mAs = milliamperesecond; SSD = source-to-skin distance; ESE = entrance skin exposure. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

181 of 203 

Table A.7.2 External Exposure Default Assumption Comparison for Y-12, Hanford and the 

Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS

Missed Photon Dose 

Application

Applies to workers with 

not recorded dose because 

(1) they were not 

monitored before 1961, 

(2) there is no recorded

dose for short periods of

time after 1961, and (3) the 

worker’s dose during a

monitoring period was

recorded as zero because 

the dosimeter response was 

less than the MDL. (Kerr 

2003, pg. 30)

Applies to workers with no

recorded dose because they 

weren’t monitored or their 

results are unavailable; and 

Workers who have a zero

recorded dose, (Fix 2004, pg. 

75). 

Applies to workers with no

recorded dose because they 

weren’t monitored or their 

results are unavailable; and 

Workers who have a zero

recorded dose (Scalsky 

2004, pg. 111). 

Missed Photon Dose 

Methodology

Multiply the MDL by the 

number of zero dose results

to estimate the 

maximum potential missed 

dose. (Kerr 2003, pg. 31) 

Divide the MDL by 2, and 

multiply by the number of zeros 

and not monitored periods (Fix

2004, pg. 75). Table 6E.6 (Fix

2004), provides potential

maximum photon dose by year. 

(1) For a claimantfavorable maximum 

potential missed dose, 

use the limit of

detection (LOD) 

multiplied by the 

number of zero doses 

(Scalsky 2004, pp. 111 

and 238)

(2) Divide the limit of

detection (LOD) 2, and 

multiply by the number 

of zeros and not

monitored periods;

(Scalsky 2004, pg. 

242), or

(3) Missed doses are added 

to measured doses and 

treated as a constant. 

IREP Dose 

Distribution Type for 

missed photon dose 

The selection of a normal 

distribution for the 

“Type” determines the

definition of Parameters #1 

and #2. For a normal 

distribution, Parameter #3 

is not used and Parameter 

#1 is the mean of the 

distribution of recorded

dose for each year of 

monitoring. (Kerr 2003, pg

41 and 43) 

Lognormal distribution with a 

geometric standard deviation of

1.52.

1

 The assessment at 

Hanford was based on the

assumption that uncertainties 

from individual sources

followed independent

lognormal distributions. For 

each uncertainty source, a 

factor is assigned reflecting bias 

(B) and a 95% uncertainty 

factor (K); the uncertainty 

factor was determined so that 

the interval obtained by

dividing and multiplying by this 

factor would include 95% of all

observations (Fix 2004, pg. 27). 

(1) When using the Limit 

of Detection (LOD)/2 

methodology, a 

lognormal distribution

with a geometric 

standard deviation of

1.52 in Parameter 2 of

the IREP input is used

(Scalsky 2004, pg. 

116). 

(2) When simply adding

the missed and 

measured dose, a 

constant is used. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

182 of 203 

Table A.7.2 External Exposure Default Assumption Comparison for Y-12, Hanford and the 

Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS

Missed Neutron Dose

Application

It is possible to calculate 

The missed neutron dose at

Y-12 using the MDLs 

because the neutron 

dosimeters were calibrated 

with neutron sources that

had energies similar to

those encountered in the 

workplace and more than 

90% of the neutrons to

which workers were 

normally exposed had 

energies greater than the 

500-keV threshold of the 

NTA film dosimeters. 

(Kerr 2003, pg. 46) 

Assign a missed neutron dose if

the individual worked in a

facility with a potential for 

neutron exposure, 

The vast majority of neutron 

dose to Hanford workers was

received at the 200 West Area

Plutonium Finishing Plant

(PFP) facilities (pg. 74.) There 

is potential for significant

missed dose in the 300 Area

plutonium laboratory (308, 309, 

324), the 100 Area reactor 

facilities (i.e., reactors, (B, D, 

F, H, DR, C, KW, KE), the 300 

Area accelerator (3754B), the

calibrations facilities (3745, 

318) and the Fast Flux Test 

Reactor (pg. 73). (Fix 2004). 

Assign a missed neutron 

dose if there is neutron 

monitoring between 1958

and 1962, if there is neutron

monitoring in 1971 or later, 

or there is indication of use

of the 17 keV calibration 

curve for interpretation of

beta/gamma film. Also

applies to those who 

worked with Cf or Cm, 

maintenance workers, those

involved in the PuAl target 

campaign, and those on

routine plutonium bioassay. 

If the recorded neutron dose 

is greater than the 

calculated dose, the 

calculated dose is used

(Neton 2003). 

Missed Neutron Dose

Methodology 

If the MDL for NTA film

is used in estimating the 

missed neutron dose, it

should be multiplied by

1.10 for workers in the 

Calibration Laboratory and

by 1.05 for workers in the 

Enriched Uranium Storage 

Area of Building 9212 and

the Nondestructive 

Analysis Laboratory. It is 

also possible to estimate 

missed neutron dose in

some facilities by use of

neutron-to-photon dose 

ratios (NIOSH 2002).

However, the only Y-12

facility where a neutron-tophoton dose ratio is 

expected to provide a 

reasonably reliable 

estimate of the missed dose 

is for workers in the 

Enriched Uranium Storage 

Area of Building 9212. 

(Kerr 2003, pg. 46-47)

A neutron-to-photon ratios is

applied to missed and recorded

photon dose for nonmonitored 

worker and workers with 

inadequate neutron monitoring. 

The upper 95% value is used

for the maximizing technique. 

The mean value is used for the 

best-fit technique (Fix 2004, pg. 

75-77). 

A neutron-to-photon ratios

is applied to missed and 

recorded photon dose for 

nonmonitored worker and 

workers with inadequate 

neutron monitoring (i.e., 

prior to 1971). The upper

95% value is used for the 

maximizing technique. The

geometric mean value is

used for the best-fit 

technique (Scalsky 2004, 

pg. 240-241). 

After 1970, the assignment

of missed dose is based on

the limit of detection

provided in Table E-10

(Scalsky 2004, pg. 241-

242). 

It appears that an ICRP 60 

correction factors is applied 

to missed dose; however,

this is unclear in the TBD 

(Scalsky 2004, pg. 110).
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

183 of 203 

Table A.7.2 External Exposure Default Assumption Comparison for Y-12, Hanford and the 

Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS

IREP Dose 

Distribution Type for 

missed neutron dose 

In order to calculate the 

neutron dose input to IREP

(see Table F-1), the 

recorded neutron dose must

be separated into neutron 

energy groups as shown in

Table F-3 and subsequently

converted into ICRP 60 

methodology (ICRP 1990).

(Kerr 2003, pg. 43)

Lognormal distribution with a 

geometric standard deviation of

1.52.

1

Lognormal distribution

with a geometric standard 

deviation of 1.52.

1

 

IREP Exposure Rate Acute for beta and photon, 

and Chronic for neutron 

Kerr 2003, pg. 41)

Acute for beta and photon 

Chronic for neutron 

(Fix 2004, pp. 8, 59, and 69, 

respectively) 

Acute for beta and photon 

Chronic for neutron 

(Scalsky 2004, pg. 87 and 

235, respectively). 

IREP Radiation Type 

(default) 

Photon, 30-250 keV p.20

Electron, > 15 keV

Neutron 0.1-2 MeV

(Kerr 2003, p. 23)

Photon, 30-250 keV

Electron, > 15 keV

Neutron, 0.1-2 MeV

(Fix 2004, pg. 29)

Photon, 30-250 keV

Electron, > 15 keV, 

Neutron, 0.1-2 MeV

(Scalsky 2004, pp. 49, 236, 

and 237, respectively)

Organ dose

conversion factor 

Once the adjusted photon 

and neutron doses have 

been calculated for each 

year, the values are used

to calculate organ doses of 

interest using the NIOSH 

External Dose

Reconstruction

Implementation 

Guideline (NIOSH 2002). 

Parameter #2 is the 

standard deviation of the

normal distribution for the

organ dose. The individual

dose result for each 

dosimeter exchange period

will be available to

calculate the mean and 

standard deviation for each 

year. If not available, the 

adjusted organ dose can be 

used for each year and a 

default standard deviation 

value used for parameter 

#2. (Kerr 2003, pg. 46)

The dose conversion factors for 

each, organ, radiation type, and 

energy ranged from OCAS-IG001 are used. If the exposure 

geometry cannot be determined, 

default values are found in

Table 6E-9 (Fix 2004, pg. 77). 

No separate value is provided

for the maximizing approach. 

For the maximizing 

approach, a value of one is 

used, pg. 61.. 

For the best-fit analysis, the 

dose conversion factors in 

the external dosimetry 

guide for the relevant

exposure geometry. 

OCAS-IG-001 Appendix A 

(NIOSH 2002) contains a 

detailed discussion of the

conversion of measured 

dose to organ dose 

equivalent, and Appendix B

contains the appropriate 

dose conversion factors 

(DCFs) for each organ, 

radiation type, and energy

range based on the type of

monitoring performed. 

(Scalsky 2004, pg. 242)

Exposure geometry Default exposure: 

Likely non-compensable 

workers – 100% AP

Compensable worker – 

50% AP, 50% ROT 

Compensable supervisor – 

Default exposure: 

Likely non-compensable 

workers - 100% AP 

Compensable worker – 50% 

AP, 50% ROT

Compensable supervisor –

Default exposure: 

Likely non-compensable 

workers - 100% AP 

Compensable worker – 

50% AP, 50% ROT 

Compensable supervisor –
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

184 of 203 

Table A.7.2 External Exposure Default Assumption Comparison for Y-12, Hanford and the 

Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS

50% AP, 50% ISO 

(Kerr 2003, pg. 34)

50% AP, 50% ISO. 

(Fix 2004, pg. 77)

50% AP, 50% ISO. 

Dose reconstructor has the 

option to choose the most 

appropriate exposure 

geometry for the individual. 

(Scalsky 2004, pg. 242)

Photon Adjustment

Factors 

(Recorded Dose) 

Multiply reported deep 

photon dose by a factor of

1.34 to estimate Hp(10). 

There is one group of Y-12

workers for which an

adjustment in the recorded

photon dose is

recommended (see Section 

6.3.4.1, pg. 19-29 ). These 

workers performed waistlevel handing jobs in DU

process areas (see Table 

6.3.4.1-1, pg. 20)). 

No adjustment for the multielement dosimeter, TLD, or 

gamma dose.

For 200 Area plutonium

workers prior to 1957, the 20% 

of the open window dose is

added to the penetrating dose 

(Fix 2004, pg. 73).

Multiply by 1.119 for years 

prior to 1987. Multiply by

1.039 for 1987. No

adjustment is needed post 

1987, (Scalsky 2004, pg. 

238). 

Note: Taylor et al. (1995)

indicates that the 1.119 

adjustment factor should be 

applied through 1985 and 

the 1.039 adjustment factor

should be applied for 1986. 

No correction is required

for 1987 and after. 

IREP Dose 

Distribution Type for 

recorded photon dose 

Constant , pg. 43 Constant.

1 Constant. The adjustment

factor encompasses the

uncertainty so no additional 

uncertainty factors are 

included. 1

Recorded Neutron 

Dose Adjustment 

Factor 

(Prior to 1971 – SRS; 

Prior to 1972 Hanford)

The missed neutron dose at

Y-12 using the MDLs 

because the neutron 

dosimeters were calibrated 

with neutron sources that

had energies similar to

those encountered in the 

workplace and more than 

90% of the neutrons to

which workers were 

normally exposed had 

energies greater than the 

500-keV threshold of the 

NTA film dosimeters. 

(Kerr 2003, pg. 46) 

NTA film is considered 

inadequate for use in dose

reconstruction due to the energy

dependence. The missed 

neutron dose approach is

applied for this period of time

(Fix 2004, pg. 48).

NTA film is considered 

inadequate for use in dose

reconstruction due to the 

energy dependence. The 

missed neutron dose 

approach is applied for this 

period of time. If the

measured dose from the 

NTA is greater than the 

calculated dose, this value 

is used and the ICRP 60

conversion factor is applied 

(Scalsky 2004, pg. 238). 

Recorded Neutron 

Dose Adjustment 

Factor (7/78-12/83) 

The recorded dose 

equivalent is a combination 

of all neutron energies. 

In order to calculate the 

neutron dose input to IREP

(see Table F-1), the 

recorded neutron dose must

be separated into neutron 

energy groups as shown in

When using of the four-chip 

HMPD during the period of its 

use from July 1978 through

December 31, 1983 in Hanford 

200 and 300 Area plutonium

facilities only, multiply the 

recorded neutron dose by 1.35. 

At all other times, divide the 

dose into the facility specific 

 In order to calculate the 

dose input for the IREP, 

Table E-1, the recorded

neutron dose must be 

separated into neutron 

energy groups as shown in

Table E-3 and subsequently

converted to ICRP 60

(1990) methodology
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

185 of 203 

Table A.7.2 External Exposure Default Assumption Comparison for Y-12, Hanford and the 

Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS

Table F-3 and subsequently

converted into ICRP 60 

methodology (ICRP 1990).

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. (Kerr 2003, pg. 43)

neutron energy bins, and 

multiply by the ICRP 60

conversion factor (Fix 2004, 

pg. 74). 

(Scalsky 2004, 235-238). 

Recorded Neutron 

Dose Adjustment 

Factor (1/72-6/78,

1/84 – present)

Divide the recorded

neutron dose into the 

facility specific neutron 

energy bins, and multiply 

by the ICRP 60 conversion

(Kerr 2003, pg. 43)

Divide the recorded neutron 

dose into the facility specific 

neutron energy bins, and 

multiply by the ICRP 60

conversion factor (Fix 2004, 

pg. 74). 

In order to calculate the 

dose input for the IREP, 

Table E-1, the recorded

neutron dose must be 

separated into neutron 

energy groups as shown in

Table E-3 and subsequently

converted to ICRP 60

(1990) methodology 

(Scalsky 2004, 235-238). 

IREP Dose 

Distribution Type for 

recorded neutron dose 

Constant Constant1 Constant The adjustment

factor encompasses the

uncertainty so no additional 

uncertainty factors are 

included. 1

Shallow Dose 

Adjustment Factors 

The information needed to

evaluate claims is directed 

to the technical parameters 

of the annual estimates of

the primary organ dose that 

is calculated from the 

dosimeter interpreted

personal dose equivalent,

Hp(10), and Hp(0.07) in

the case of skin, testicular, 

and breast cancer. These

are used as a consistent 

basis of comparison for all 

years of Y-12 occupational 

external dose starting in

1950. 

Shallow dose adjustments 

factors are not addressed in the 

TBD. 

Shallow dose adjustments 

factors are not addressed in

the TBD or SRS TIBs. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

186 of 203 

Table A.7.2 External Exposure Default Assumption Comparison for Y-12, Hanford and the 

Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS

Low-energy photons 

(< 30 keV) 

The TBD mentions no 

adjustment for low-energy

photons.

The stated Hanford practice to 

include 1/5 of the shallow dose 

based on a 16-keV calibration 

to the deep dose for Hanford 

plutonium facilities workers 

could resolve this source of

potential under-response around

17 keV (Fix 2004, pg 26). For 

200 Area workers prior to 1957, 

the 20% of the open window 

dose is added to the penetrating

dose, (pg. 14). 

1954–1981 Subtract the 

reported deep dose from the 

shallow dose for plutonium

workers. 

1982-present. Plutonium

workers are those 

individuals that worked in

321M, 221H – B line, 221F

– B line, 772F, 235F, 773A, 

736A, and other plutonium

storage areas (Neton 2004). 

(For testicular, breast, or 

skin cancer) 

IREP Dose 

Distribution Type for 

recorded shallow dose 

Not Included in the TBD. Not included in the TBD. Shallow dose is addressed 

from a technical

perspective in the TBD, but 

no direction is provided to 

the dose reconstructor 

(Scalsky 2004, pg. 97). 

IREP Radiation Type 

for recorded dose 

The primary IREP screen

used to input dose 

parameters is illustrated in

Table F-1. Assume 30-250 

KeV with a normal when 

entering distribution 

parameters. The input to 

these fields is obtained 

from the Y-12 dose of 

record. Kerr 2003, pg. 41. 

Specific to the particular

facility for beta, photon, and

neutron dose. For example, in

the reactor area 100% of the 

beta doses is assumed to be >15 

keV, 75% of the photon dose is

>250 keV, and 25% of the 

photon dose is 30-250 keV 

(Fix 2004, pg. 29).

Specific to the particular

facility for beta, photon,

and neutron dose. For 

example, in the reactor area 

100% of the beta doses is

assumed to be >15 keV,

50% of the photon dose is

>250 keV, and 50% of the 

photon dose is 30-250 keV 

(Scalsky 2004, pg. 98). 1

 These parameters were obtained from review of several dose reconstruction IREP input sheets. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

187 of 203 

Table A.7.3 Internal Exposure Default Assumption Comparison for Y-12, Hanford and the 

Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS 

Particles Size 

(default) 

5 micron 

(Rich 2005, pgs 8 & 10, and 11)

5 micron 

(Bihl 2004, pg. D-10)

5 micron 

(Scalsky 2004, Section 4.0, 

Attachment D) 

Intake Type

(default) 

Chronic prior to stand-down, the 

Y-12 program default modeling

assumption was Class Q (90% 

Super W, 10% Y) 8 um AMAD – 

pg. 41. However, during a 1994 to

1998 stand down period, it is 

assumed to be acute occurring at 

the midpoint of a quarterly

sampling frequency – pg. 12

(Rich and Chew 2005)

Chronic 

(Bihl 2004, pg. 7-9)

Chronic 

(Scalsky 2004, Section 4.0, 

Attachment D) 

Default Excretion 

Volume

Nominal 1.4 L/day – pg. 21

(Rich and Chew 2005)

Uses a urinary excretions 

value of 0.2 ug/d for

elemental analyses, 0.15 

dpm/d for 234U and 238U 

and essentially anything

detected for 235U 

(Bihl 2004, pg., 27)

1.4 liters/day 

(Volumes less than 1.4 

liters/day are corrected by

normalizing the actual 

volume to 1.4 liters/day. 

Samples recorded as activity 

per 1.5 liters are not 

corrected.) (Scalsky 2004, 

pg. 70) 

Solubility Class 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 may be based on

the monitoring data and/or

claimant favorable assumptions. – 

pg. 10. 

Uranium compounds handled at

Y-12 range from highly soluble to

very insoluble- pg. 6. 

For a workplace as varied as Y-12, 

it is clear that no single solubility 

or particle size will apply to all 

workers – pg. 9 Therefore,

claimant favorable assumptions 

should be made regarding 

solubility and uranium activity 

ratios - pg.25 

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 – pg. 43 (Rich and

Chew 2005)

For the maximizing 

approach, the most 

claimant favorable 

solubility type for the 

organ of interest is used. 

For the best-fit approach

the most appropriate 

solubility type can be

used. Inhalation class and 

lung absorption type for 

uranium is found in Bihl

2004, Table 5.2.5-3, pg.

24 ).

For the maximizing 

approach, the most claimant 

favorable solubility type for 

the organ of interest is used. 

For the best-fit approach the 

most appropriate solubility 

type can be used (Scalsky

2004, pg. 85). 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

188 of 203 

Table A.7.3 Internal Exposure Default Assumption Comparison for Y-12, Hanford and the 

Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS 

Intake Date for 

Hypothetical Intake

(excluding tritium)

TBD does not specify. First day of employment

or the first day of

operation of the facility 

where the worker was

assigned. For separation 

plants, chronic intakes 

would apply from either 

the first day of work for 

the worker or the start up 

of the plant, December 

1944 for T Plant and 

April 1945 for B Plant 

(Bihl 2004, pg. 8). 

Acute inhalation on January 1

in the first year of 

employment (Scalsky 2004, 

pg. 85; Brackett 2003, pg. 3). 

Tritium Missed

Dose Application

TBD does not specify. Assigned to workers who

worked in 108-B, the 300 

Area Test Reactors, and 

in some cases where 

work location was 

unknown or variable. 

Those who never wore a 

dosimeter and had no

bioassay results were 

assigned environmental

doses (Bihl 2004, pgs. 

21–22). 

Assigned to workers 

monitored for external dose,

but having no bioassay. For 

workers not in the dosimetry 

or bioassay monitoring

program, the missed internal 

dose is based on

environmental intake only.

Scalsky 2004, pg. 84; Duncan

2003, pgs 6 and 12) 

Basis for Tritium

Missed Dose 

Not included in the TBD Tritium urinalysis was 

not perfected until 1961. 

Liquid scintillation

counting for tritium

likely was started in 1958 

(Bihl 2004, pg. 21-22). 

From 1949 to 1960 the 

MDA was 5 uCi/L and 

from 1961 to 1981 the 

MDA as 1 uCi/L. Later 

in 1982 the MDA 

changed to 10 dpm/ml

and in 1991 to 20

dpm/ml, (Bihl 2004, pg

22). Tritium intakes 

were accounted for as

part of external dose until 

about 1986-87 (TBD 

does not explain 

methodology), when they

were entered in the dose

database as internal dose 

(Bihl 2004, pg. 12 & 22). 

Dose calculated based on the 

tritium reporting level for a 

particular time period

(Scalsky 2004, pg. 67; 

Duncan 2003, pg. 6). 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

189 of 203 

Table A.7.3 Internal Exposure Default Assumption Comparison for Y-12, Hanford and the 

Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS 

Hypothetical Intake

Application

Hypothetical intake applications

not discussed in TBD 

Applied to individuals 

who wore a dosimeter 

but did not have any 

bioassay (Bihl 2004, 

pg. 48). 

Applied to claims with nonmetabolic and digestive tract 

cancers (Scalsky 2004, pg. 

85; Bracket 2003, pg 2). 

Basis for missed 

internal dose from

radionuclides other 

than tritium

(1) From 1950 to 1989, collection

of routine urine samples for 

fluorometric analysis of uranium

was initiated – pg. 23. 

(2) The internal dosimetry

program has included limited 

monitoring for Cs-137, Tc-99,

thorium, plutonium, Ac-228, and 

tritium, among other 

radionuclides. 

 (Rich and Chew 2005, pg. 43)

(1) Individuals with no

external or internal 

monitoring data were 

assigned an

environmental

internal dose, (Bihl 

2004, pg. 48)

(2) For those individuals

with external 

monitoring but no or

limited internal

monitoring, the 

approach was year 

dependent. For 1947

through 1952, daily 

intakes at 10% of the 

respiratory

protection required 

value for 40

hours/week were 

assumed. Iodine was

assumed to be at 0.1 

times the vapor 

index. For 1953

through 1988, daily 

intakes were based

on an exposure to 

airborne 

concentrations at 

10% of the limiting 

air concentration for 

four hours per week, 

(Bihl 2004, pg. 49). 

(3) From 1989 through

the present, a daily 

exposure at 5% of

the limiting air 

concentration for 4 

hours per week was 

assumed, (Bihl 2004, 

pg. 50). 

(4) For monitored

workers with no

confirmed intake, a 

maximum intake is 

(1) Individuals with no

external or internal 

monitoring data were 

assigned an

environmental internal 

dose (Scalsky 2004, pg. 

84; Bracket 2003, pg. 2). 

(2) For those individuals

with external monitoring

but no or limited internal

monitoring, an annual

missed tritium dose and 

environmental dose from

uranium, plutonium and 131I are assigned as

internal dose. It is also 

reasonable to pick a 

fission or activation

product that produces the 

largest dose to the organ 

of interest dose (Scalsky 

2004, pg. 84; Bracket 

2003, pg. 8). 

(3) Highest five intakes for 

various nuclides are 

applied to those 

individuals with nonmetabolic or digestive 

system cancers (Bracket 

2003, pg. 2). 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

190 of 203 

Table A.7.3 Internal Exposure Default Assumption Comparison for Y-12, Hanford and the 

Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS 

determined by using 

the MDA of the last 

sample as an upper 

bound (Bihl 2004, 

pg. 47). 

Radionuclides

included in the

Hypothetical Intake

Not discussed in the TBD. Variable by facility and 

organ of interest. Alpha 

intakes are assigned for 

the Plutonium Finishing

Plant (PFP), the 200 Area 

Fuel Separations Plants, 

U-Plant, C-Plant, the 300

Area Fuel Fabrication 

Facilities 209E, 120, 324, 

325, 327, the tank farms 

and evaporator facilities 

(0.5 times the alpha

intake), and where work

location in unknown or 

highly variable. Alpha 

intakes are based 

primarily on 234

U or 239Pu. Beta/gamma 

intakes are assigned for 

all facilities except PFP, 

209E, 120, the 300 Area 

Fuel Fabrication 

Facilities, 108-B, and UPlant. Tritium intakes 

are assigned for the 108-

B Building, the 300 Area 

Test Reactors, and in 

some situations where 

work locations are 

unknown or variable. 

The particular 

beta/gamma radionuclide

and its solubility class are 

determined based on the 

organ of concern. For 

some facilities and 

periods of time it is 

specified (Bihl 2004, pp. 

51-52). 

241Am/241Pu (M), 244Cm (M), 60Co (S), 137Cs (F), 237Np

(M), 238Pu (M), 239Pu (M), 90Sr (F), 234U (F), and 238U 

(F). (Bracket 2003, pg. 9)

Default Activity 

Ratios Pu Mixture 

Due to the uncertainty in the U233 to U-232 ratio process, 

claimant-favorable assumptions 

should made about solubility and 

uranium activity ratios. 

(Rich 2005 pg. 25) 

Not specified in the TBD. Ten-year old 12% plutonium

mix (Scalsky 2004, pg. 66). 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

191 of 203 

Table A.7.3 Internal Exposure Default Assumption Comparison for Y-12, Hanford and the 

Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS 

Activity Fractions

for other Mixtures 

Not discussed in the TBD. Activity fractions are 

provided for uranium

mixtures, Table 5.2.5-3,

pg. 24, weapons and fuel 

grade plutonium, Table 

5.2.1-3 pg. 16, and

recycled uranium

impurities. Table 5.2.5-

2, pg. 24. Default 

mixtures based fission

product urinalysis was 

developed by time period

and organ of concern

(Bihl 2004, pg. 10, 

Attachment D). 

Activity fractions are facility 

dependent. The activity 

fractions are taken from the 

Internal Dosimetry Technical 

Basis Manual (WSRC 1990). 

The information for these 

ratios was obtained from

safety analysis reports,

personal interviews, open 

literature, etc. 

Radionuclides of

Concern for 

Monitored Workers 

In addition to 234U, 235U, and

238U, the following radionuclides 

are identified in the TBD: 3

H, 90Sr, 99Tc 

• 228Th, 232Th, 

• 232U, 233U, 236U 

• 238Pu, 239Pu, 241Pu 

• 237Np, 241Am 60Co and 95Zr/

95Nb for 

organizations outside Y-12, 

(Cofield 1961) 

• 210Po 

• 40K, 137Cs 

(Rich and Chew 2005, pg. 12)

Radionuclides of concern 

were based on the in vivo

and in vitro bioassay data 

of the individual, or the

minimum detectable 

activity for a particular 

radionuclide. 

Radionuclide

assumptions varied by

facility and organ of

interest (Bihl 2004, pg. 

13). 

Radionuclides of concern 

were based on the in vivo and

in vitro bioassay data of the 

individual (Scalsky 2004,

pgs. 66 & 67). Although the

TBD provides activity 

fractions in Attachment A, it 

is not clear how these activity 

fractions are used in dose

calculations. 

Tritium Dose for 

Monitored Workers 

As of 1957, personnel engaged in

processing materials with a

potential for tritium contamination

submitted three urine samples per 

month – (Rich 2005 pg. 26) 

Tritium urinalysis was 

not perfected until 1961. 

Liquid scintillation

counting for tritium

likely was started in 1958 

(pg. 21-22). From 1949 

to 1960 the MDA was 

5 uCi/L and from 1961 to

1981 the MDA as 

1 uCi/L. Later in 1982

the MDA changed to

10 dpm/ml and in 1991

to 20 dpm/ml (pg. 22). 

Tritium intakes were 

accounted for as part of 

external dose until about 

1986–1987 (TBD doses 

not explain 

methodology), when they

were entered in the dose

database as internal dose 

Based on the reporting level

if the tritium bioassay is less 

than this level, or the actual 

bioassay result if it is greater 

than the reporting level. 

Organically Bound Tritium

and Stable Metal Tritides are 

not considered (Brackett 

2003, pg. 6). 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

192 of 203 

Table A.7.3 Internal Exposure Default Assumption Comparison for Y-12, Hanford and the 

Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS 

(pp. 12 & 22). (Bihl 

2004, pg. 12 & 22)

Internal Dose for 

radionuclides other 

than tritium

The primary focus on internal 

dose control has been on uranium

compounds and alloys over a wide 

range of 235U enrichment. 

At times, the uranium compounds 

processed may have contained 

impurities with radiological health

implications. 

Pg. 7 

Uranium has always been the 

dominant contributor to collective 

internal dose at Y-12. Monitoring

for other radionuclides has been 

performed on a limited basis – pg. 

16 

Analyses for other radionuclides 

were performed on an as needed

basis - pg. 20 

(Rich and Chew 2005)

Based on either actual

bioassay values for 

positive values. Based

on a chronic intake over

the entire exposure 

period with the last 

sample assumed to be at 

the MDA (Bihl 2004, pg. 

47). 

Based on either actual

bioassay values or detection

levels for bioassay

techniques. For nonmetabolic cancers, the 

maximizing approach is used

(Scalsky 2003, pg. 85). 

Basis for prebioassay program

doses 

Urine by fluorometry (usually for 

an individual who worked in

areas with natural/depleted

uranium; prior to 1950, also used

for workers in enriched uranium

areas - pg. 11 and 21. 

Plutonium in Urine by gross alpha 

counting.

Was used before 1988 – pg. 22

(Rich and Chew 2005)

Air concentration 

tolerance or limits. 

(Bihl 2004, pg, 7)

Not included in the TBD. 

Ingestion Not included in the TBD. Assigned during periods

were air sampling was

used to determine 

internal dose. The 

quantity is based on the 

air concentration level or

on the guidance provided

in Estimation of 

Ingestion Intakes

(NIOSH 2004). 

(Bihl 2004, pg. 8) 

Not included in the TBD. 

Surrogate

Radionuclide in

IMBA for 65Zn/95Zr 

Not included in the TBD. Not included in the TBD. 137Cs used as a surrogate. 

Surrogate Adjustment factor

= 2.43. 

(Brackett 2003, pg. 9)

Surrogate

Radionuclide in

IMBA for 106Ru/

144Ce/95Nb 

Not included in the TBD. Not included in the TBD. Radionuclides not available 

in IMBA. 90Sr used as a 

surrogate. Surrogate 

Adjustment factor = 7.25 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

193 of 203 

Table A.7.3 Internal Exposure Default Assumption Comparison for Y-12, Hanford and the 

Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS 

(Brackett 2003, pg. 9). 

Surrogate

Radionuclide in

IMBA for 242Cm/

252Cf 

Not included in the TBD. Not included in the TBD. Radionuclides not available 

in IMBA. 244Cm used as a 

surrogate. Surrogate 

Adjustment factor = 1.09 

(Brackett 2003, pg. 9). 

IREP Radiation 

Types for 

Hypothetical Intake

Not included in the TBD Alpha

1

Beta: >15 keV1

Photon: > 250 keV1

Tritium: < 15 keV1

Alpha 

Beta: >15 keV 

Tritium: < 15 keV 

(Bracket 2003, pgs. 8 & 12)

IREP Dose 

Distribution Type

Not included in the TBD Constant

1 Constant (Brackett 2003, pg. 

12) 

Internal Dose 

Uncertainty 

The use of the rate method for 

determination of NU and DU to

estimate daily urine excretion, and

hence radionuclide elimination, 

contributed to the uncertainty

associated with any given 

measurement. 

(Rich 2005 pg. 20) 

For the missed dose 

assignments, the value 

entered includes the 

uncertainty. For dose 

assignments based on

monitoring data, the 

following values can be

applied as a standard

deviation:

(1) 0.3 times the MDA 

or reporting level, or 

(2) 0.5 times the MDA 

for chest counting. 

Actually report errors can 

be used if available (Bihl

2004, pg. 46). For air

concentration data, a 

triangular distribution 

with zero as the 

minimum, the derived

values as the mode, and

twice the mode as the

maximum is used (Bihl 

2004, pg. 7). 

For the missed dose 

assignments, the value 

entered includes the 

uncertainty. 1 No direction is 

provided to the dose 

reconstructor for dose 

assignments based on

monitoring data. 

Other Comments None. Informs the dose 

reconstructor of limited

use radionuclides such as 14C, 232Th, radon, 90Y, 227Th, 227Ac, and 32P 

(Bihl 2004, pg. 32)

None. 

1

 These parameters were obtained from review of several Hanford dose reconstruction IREP input sheets. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

194 of 203 

Table A.7.4 Environmental Exposure Default Assumption Comparison for Y-12, Hanford 

and the Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS 

Application The data from the scoping

survey should be used to

calculate an external dose for 

unmonitored workers (Foley

and Carrier 1990). The dose

rates shown in Table D-7 

represent the external 

exposure rates measured site 

wide. The exposure rates were

converted to dose equivalent

rates assuming a quality factor 

of one (1). 

(Ijaz and Adler 2004, pg. 45)

Environmental doses are 

assigned to personnel with no

bioassay and no evidence of

having worn a dosimeter at

the Hanford Site (Bihl 2004, 

pg. 48). 

Apply the annual internal 

and external environmental 

dose for each full or partial 

year of employment for the

maximizing approach. 

Dose reconstructors are 

instructed to use only the 

maximum annual intakes in 

Table C-17 for the 

maximizing approach

(Scalsky 2004, pg. 179). 

For the best-fit approach, 

modifications can be made 

for partial year of

employment. No

environmental dose is

assigned if the background 

is not subtracted from the 

workers badge (Scalsky 

2004, pg. 62). 

Sources of 

Environmental

Releases 

Considered

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, 

2. Exposures received from

radiation levels emanating

from buildings and storage 

areas. 

(Ijaz and Adler 2004, pg. 45)

T-plant particles and iodine, 

B-Plant particles and iodine, 

REDOX particles and iodine, 

PUREX particles and iodine, 

Z-Plant particles, reactor 

noble gases, and tritium from

108B Building

(Savignac 2003, pg. 18). 

The TBD heavily 

references the Cummins 

(1991) and CDC (2001)

documents, and dose not

include many of the base 

assumptions from those 

reports in the TBD. It is 

apparent that releases from

the reactors and separations

areas were considered. 

Source Term Basis Aerial radiological surveys and 

flyovers for collimated sources 

(pg. 28); waste disposal area 

report (pg 28); outdoor

radiological and chemical 

scoping surveys (pg. 29); and 

data for Y-12 monitoring

stations (pg. 38) (Ijaz and

Adler 2004) 

Hanford Works 

environmental reports; 

Methods for Estimating 

Radiation Doses from ShortLived Gaseous Radionuclides 

and Radioactive Particles 

Released to the Atmosphere 

During Early Operations at 

Hanford (Till et al. 2002). 

Radioactive Releases from 

the Savannah River Plant 

1954–1989 (Cummins

1991), Savannah River Site 

Dose Reconstruction 

Project Phase II: Source

Term Calculation and

Ingestion Pathway Data

Retrieval, Evaluation of

Materials Released from 

the Savannah River Site

(CDC 2001), SRS 

meteorology data, SRS 

environmental reports for

1993-2001. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

195 of 203 

Table A.7.4 Environmental Exposure Default Assumption Comparison for Y-12, Hanford 

and the Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS 

Methodology Due to the limitations of the 

available data and the 

complexity of modeling 

releases, an alternative

approach was developed for 

estimating airborne uranium

concentrations. An

empirical relationship was 

developed using on site

measured air concentrations

and estimated uranium release 

estimates. This approach

circumvents the need for air

dispersion modeling by

providing a direct relationship

between uranium air 

concentrations and uranium

releases. 

(Ijaz and Adler 2004, pg 14)

Puff advection (RATCHET) 

model (Savignac 2003, pg. 

14) 

Gaussian model (Scalsky

2004, Section 3.1.1) 

Type of Releases Concentrations of airborne 

uranium within the Y-12

perimeter were measured from

1983 through 1999. This 

seventeen-year period

represents less than 30% of the 

operational history of the site.

In addition, these data were not 

collected during the time

operations and releases were at

their maximum. Therefore 

these data cannot be used as 

estimators of historic air 

concentrations. Prior to 1983, 

the various networks of air

monitoring stations in and 

around the ORR were

established to monitor for the 

FP releases. 

(Ijaz and Adler 2004, pg. 13)

Calculations included routine 

and identified non-routine 

releases. Estimates include 

inhalation of radionuclides in

air, direct external radiation 

from plumes, and physical 

contact with particulate 

radionuclides on skin. 

The TBD heavily 

references the Cummins 

(1991) and CDC (2001)

documents, and dose not

include many of the base 

assumptions from those 

reports in the TBD. 

Ventilation Rate 

(m3/year) 

2,400 (default); pg. 27 

Based on an exposure 

assumption of 2000 hours/year 

and an inhalation rate of

1.2 m3 hour. 

(Ijaz and Adler 2004, pg. 38)

2,400 (default);

Based on 1.2 m3

/hour ±

0.4 m3

/hour 

(Savignac 2003, pg. 16)

2,400 (default);

Adjustments can be made 

for light and heavy work

(Scalsky 2004, pg. 162). 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

196 of 203 

Table A.7.4 Environmental Exposure Default Assumption Comparison for Y-12, Hanford 

and the Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS 

Exposure Time

(hours/week) 

40 40 (Savignac 2003 pg. 24) 40 with a 1.25 conversion

factor to increase the 

exposure time to 50 hours 

per week (Scalsky 2004, 

pg. 61). 

Mobile Workforce Information not included in the 

TBD. 

Information not included in

the TBD. 

Assign the maximum dose

listed for any area onsite. 

Facility Specific 

Workforce 

Information not included in the 

TBD. 

Information not included in

the TBD. 

Assign the maximum dose

listed for any area onsite for 

the maximizing approach. 

Assign an area specific 

environmental dose based

on the work location of the 

worker for the best-fit 

approach (Scalsky 2004, 

pg. 61). 

Radionuclides

Considered for

External Dose

234/235U, 238U – pg. 27 

Aerial surveys identified 

specific radiological sources

including 60Co, 234mPa and

137Cs. – pg. 28

(Ijaz and Adler 2004)

)41Ar,131I, 106Ru (Savignac 

2003, pgs. 19 and 23) 

41Ar, (Scalsky 2004, pg. 60)

Radionuclides

Considered for

Submersion Dose 

41

Ar (Savignac 2003, pg. 17) 41Ar, (Scalsky 2004, pg. 59)

Submersion DCF Information not included in the 

TBD. 

Federal Guidance Report No. 

13, Cancer Risk Coefficients 

for Environmental Exposure 

to Radionuclides, 1999. 

Assumed values from the 

Federal Guidance Report 

12 (EPA 1993). 

(Scalsky 2004, pg. 60)

Radionuclides

Considered for

Internal Dose. 

Occupational Internal Dose 

TBD has yet to be released. 

3

H,131I-131mXe, 144Ce-144Pr, 137Cs-137Ba, 239Pu, 103Ru- 103mRh, 106Ru106Rh, 90Sr-90Y, 95Zr-95Nb (Savignac 2003, pg. 

8) 

3

H,131I, 238Pu, 239Pu, 240Pu, 234U, 235U, and 238U 

(Scalsky 2004, pg. 51)

Soil Outdoor radiological and

chemical scoping surveys 

provided measurements of both 

gamma ray exposure rates and 

the collection of

surface soil samples. Specific 

parameters are not specified in

the TBD. 

(Ijaz and Adler 2004, pg. 6-7)

Not included in the TBD. Density = 1,600 kg/m3

Surface Factor = 0.08 

Resuspension Factor =1E9/m 

(Scalsky 2004, pg. 59)

Liquid Effluents Not included in the TBD. Not included in the TBD. Not included in the TBD. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

197 of 203 

Table A.7.4 Environmental Exposure Default Assumption Comparison for Y-12, Hanford 

and the Savannah River Site

Description of 

Assumption 

Y-12 Oak Ridge Hanford SRS 

Organ Dose

Conversion Factor

Not included in the TBD. Not included in the TBD. 1.0 is used in the 

maximizing approach. The

organ dose conversion

factors in the external 

dosimetry guide for the 

relevant exposure geometry

are used in the best-fit 

analysis (Scalsky 2004, pg. 

61). 

IREP Rate Chronic Chronic

1

 Chronic (Scalsky 2004, pg. 

61) 

IREP Radiation 

Type 

Photon, 30-250 keV Photon, 30-250 keV1 Photon, 30-250 keV 41Ar , 100% photon, > 250

keV (Scalsky 2004, pp. 60

&61) 

IREP Dose 

Distribution Type

Constant Constant. Doses and intake 

quantities provided with a 

geometric mean and standard 

deviation. There is no

direction on how these values 

should be entered into IREP. 

Constant. Doses and intake 

quantities provided with a 

50th-percentile and a 

geometric standard 

deviation. A 95th percentile 

for the source term is 

estimated as 25% greater

than the 50th percentile 

(Scalsky 2004, pg. 60). 

Special 

Considerations for 

Uranium and

Plutonium

Non mentioned in the TBD. Not applicable. The isotope yielding the 

maximum organ dose was 

assumed at 100% rather

than applying a mixture 

(Scalsky 2004, pg. 59). 

Other Aerial surveys identified

specific radiological sources

including 60Co, 234mPa and

137Cs. 

Other sources included 

collimated sources and x-ray 

machines located in buildings. 

Multiple flyovers indicate that 

these sources were not in 

constant operation. 

(Ijaz and Adler 2004) 

The four chemical 

separations plants, T Pant, B

Plant, REDOX Plant and the 

PUREX plant, along with the 

plutonium handling Z plant

are shown in Figure 4.1.1 to

be the most important release 

points at Hanford (Savignac 

2003). 

1955 values are assigned to 

1952, 1953, and 1954

(Scalsky 2004, pg 54)

1

 These parameters were obtained from review of several Hanford dose reconstruction IREP input sheets. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

198 of 203 

ATTACHMENT 8: URANIUM PROCESS FLOW AT THE Y-12 PLANT 

(DOE 1985) 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

199 of 203 

ATTACHMENT 9: EDITORIAL ERRORS IN THE TBD 

ORAUT-TKBS-0014-3 

Best Estimate” Doses in Attachment 3C Tables Are in Error 

There are errors in the Attachment 3C Tables. Guidance provided in Attachment 3C (Murray 

2004) would assign an organ dose for “best estimate” that is 30% higher than the “maximized” 

organ dose. The occupational medical TBD (Murray 2004) lacks adequate development of 

medical x-ray technical factors that were subsequently provided in ORAUT-OTIB-0006 

(Kathren 2003). This leaves the TBD sections without an adequate discussion of these factors. 

NIOSH has not provided guidance for determination of x-ray dose in claimant cases involving 

cancer on the skin of the face. X-ray dose is usually determined by the exit dose to the skin of 

the back. The dose potential from photofluorography chest x-rays during the early years was 

based on an National Council on Radiation Protection, NCRP Report 102 (NCRP 1989) which 

may not be representative of the photofluorography unit at Y-12.

ORAUT-TKBS-0014-3 identifies the fact that, at Y-12, four different x-ray units were used 

between 1943 and the present for performing standard chest x-ray procedures. Correspondingly, 

Attachment 3C of the TBD provides four separate tables (i.e., Tables 3C-2, 3C-3, 3C-4, and 3C5) for organ doses associated with chest x-rays. 

• Section 3C.3.1, Assignment of Organ Doses from X-ray Procedures: Maximizing

Approach for Dose Reconstruction, provides the following guidance for maximizing dose 

estimates: 

Organ doses from X-ray procedures have been calculated and are presented in Tables 3C-2, 3, 4, 

and 5. The organ doses assigned for each X-ray procedure are the highest doses to any organ in 

the relevant group as listed in Tables 3C-2 to 5. [Emphasis added.]

Thus, it must be assumed that organ doses cited in Tables 3C-2 to 5 have already been 

‘maximized.” However, this assumption conflicts with Section 3C.2.3 that follows, as explained 

below. 

• Section 3C.2.3, Assignment of Organ Doses from X-ray Procedures: Best Estimate 

Approach for Dose Reconstruction, provides the following guidance for assigning a best 

estimate of organ dose: 

For the dose calculation, a normal distribution is applied with a standard deviation of 30%. The 

value of the standard deviation is equal to the mean value times 30%. Thus, the dose 

reconstructor should multiply the organ doses listed in Tables C-2, 3, 4, and 5 by 1.3. 

In summary, the guidance provided in Attachment 3C of the TBD would assign an organ dose 

for “best estimate” that is 30% higher than the “maximized” organ dose. This is obviously an 

error! 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

200 of 203 

ORAUT-TKBS-0014-6 

Neutron Monitoring Methods and Records

Page 11, Table 6.3.1-1, Row 2 under neutron dosimeters, shows that from 1980–1989 the NTA 

film was used for fast neutrons and TLND dosimeters for other energy neutrons. This appears to 

be in contradiction to that stated on Page 16 for the period of 1986–1989 where TLND only was 

used for neutron dose determination. This apparent contradiction needs to be corrected. 

In Table 6.3.1.1, Page 11, of the TBD, NTA film is listed as being used from 1950–1980, but in 

Table 6.3.1.2, Page 12, NTA film is listed as being used from 1948–1960 and in Table 6.3.2.1, 

Page 14, it list the start of NTA film use in 1949. This apparent conflict needs to be corrected. 

Page 19, Table 6.3.3.2-1, states that the Angular response of Y-12 neutron dosimeters is likely 

too low because of its lower response at non A-P angles (the text before this table discusses both 

NTA film and TLNDs). However, Table 6.3.4.3-1, Page 26 states that the NTA film response is 

likely too high because of its increased response at other than A-P angles, and that the TLD 

response would decrease at other than A-P angles. This conflict needs to be resolved or clarified. 

Film Badge Energy Response

Page 15, Figure 6.3.2.1-1 shows the two-element film badge to over respond in the 77 keV to 

200 keV photon energy range as compared to Hp(10). However, the first paragraph on Page 16, 

and Table 6.3.2.1-1, states the range is 118 keV to 208 keV. 

Table 6.4.2.1.1 Footnote

Page 30, Table 6.4.2.1-1, footnote b, list the figure number as 4.6.2-1; it should be Figure 

6.4.2.1-1. 

Table 6.4.2.2.1 ICRP 60 Correction Factors 

Page 30, Table 6.4.2.2-1, and the text above it, uses the term “ICRP 60 correction factors”. 

Actually, these factors appear to be derived from site specific terms, using ICRP 60 (ICRP 

1990). They are not directly from ICRP 60. Rewording this term would make it more 

applicable. 

ORAUT-TKBS-0014-6 (Kerr 2005, pp. 28 & 43) states that the dose should be adjusted by a 

factor of 1.34 for workers in the DU process waist-level handling operators prior to 1991. The 

TBD, however, does not specify the starting date for the use of this adjustment in photon dose. 

Clarification is needed concerning the application of this dose adjustment factor. 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

201 of 203 

ATTACHMENT 10: URANIUM SEPARATION PROCESS (1943–1947) 

The TBD does not provide a complete historical account for dose reconstruction during this 

period. This is illustrated by: 

(a) The number of workers involved is not discussed and is quite large; 

(b) Levels of uranium in air were extraordinarily high at the Calutron operations; 

(c) There is little if any evidence of respiratory protection; 

(d) Individual worker dosimetry was virtually non-existent; and 

(e) By the late 1940s, AEC officials and advisors were made aware that uranium

workers were being excessively exposed to airborne levels of uranium well above 

the default limits meant to protect against radiation set in World War II.. 

Between 1943 and 1947, the Tennessee Eastman employed approximately 40,000 workers, 

including a large number of women ((Herken 2002)) to operate the Calutrons and to recover and 

process uranium for weapons in Y-12 buildings 9201-1, 9201-2, 9201-3, 9201-4, 9202, 9203, 

9204-1, 9204-2, 9204-3 ,9204-4, 9206, 9207, 9211, and 9212 (Polednak 1982). 

By any scale, the operation there was mammoth. Plans called for installing a pair 

of 500-tank Calutron race tracks end-to-end in twin two-story buildings, each 

measuring four football fields long. The racetracks were on the second story; 

pumps and plumbing for the vacuum system occupied the first floor. Logistic and 

personnel requirements were in proportion. Every pair of vacuum tanks required 

an individual operator seated at a console, continually adjusting the current to 

focus on the beam. An army of technicians was needed to monitor the orange 

uranium-oxide feed material for the beam and later scrape the errant green 

"gunk" -- uranium salts dissolved in carbon tetrachloride -- from the insides of 

each tank. An army of chemists would separate out the silvery white 

powder containing uranium-235 that was left in the receivers following each 

week-long run (Herken 2002).

By 1948, the Atomic Energy Commission’s (AEC) manager of the Oak Ridge site “submitted a 

report on radiation history of employees,” which was reviewed by the contractor and AEC 

representatives, including the AEC’s Advisory Committee for Biology and Medicine (ACBM). 

(AEC 1948) The report recommended: 

(1) a terminating employee be provided with a statement that he has not exceeded 

the permissible exposure to radiation, or if he has exceeded the permissible 

exposure he be made aware of this fact by the physician giving the exit 

interview; [Emphasis added] (2) that there be a clearer policy on release of 

information on radiation exposure records and other medical records to the 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

202 of 203 

contractor’s insurance and life insurance companies; (3) that the terminating 

employee be advised that if he is to work with radiation in the future his new 

employer can make arrangements to procure his past radiation exposure history; 

(4) that a group of qualified radiologists and physicians be available for 

consultation by any person who feels that he has been damaged by radiation at 

an AEC installation. [emphasis added] (AEC 1948) 

According to the minutes of the meeting of the ACBM in December 1948, 

...a full discussion ensued and each item was considered in detail. It was the 

opinion of the Committee that in place of the stated recommendation a 

terminating employee should be advised at the exit interview as to the care that 

the A.E.C. utilizes in protecting each employee. [emphasis added] (AEC 1948) 

Subsequently, in June 1949, the ACBM Committee was given a presentation given by Dr. 

Sterner regarding exposure to workers in uranium processing plants. (AEC 1949) According to 

the presentation: 

The original levels that were set during the earlier days were 500 micro-grams 

per cubic meter for insoluble uranium dust and 150 micro-grams per cubic meter 

for soluble. We considered a hazard at that time a chemical hazard and not a 

radiation hazard, [emphasis added] however on reconsideration there is now an 

opinion put forth that even as low as 50 micro-gram, per cubic meter, there is a 

questionable hazard from long-term alpha emitters. 50 micrograms per cubic 

meter is equivalent to about 70 alpha- counts per minute, per cubic feet and that 

is the pinch mark present at which we are striking for control in these plants. 

(AEC 1949) 

Sterner also informed that uranium processing workers were being excessively exposed –some at 

levels 125 times greater than the default standard adopted in World War II. (AEC 1949). 

By 1982, a retrospective dose assessment done by ORAU of workers at the Tennessee Eastman 

Company (TEC) Y-12 facility was published. (Polednak 1982) The study excluded all female 

workers. The authors reported that certain operations, particularly recovery areas, involved 

“concentrations as high as 9,000 ug/m3

” (Polednak 1982). It was not definitively determined if 

workers wore respirators. Moreover, “Job titles (such as ‘operator’) were not precise enough to 

identify workers with highest exposure” (Polednak 1982). 

The authors assumed that all alpha chemical workers were exposed to an average concentration 

of insoluble uranium compounds of 150 ug/m3

. (Polednak 1982) On this basis, workers 

employed for one year during this period “could have received lung doses of about 6.3 rad or 130 

rem in the 130-year period” (Polednak 1982). 
Effective Date:

September 19, 2005

Revision No. 

0 

Document No.

SCA-TR-TASK1-0007 

Page No. 

203 of 203 

Although uranium dust concentrations were found to be lower in the beta chemistry area, 

enrichment levels were at least 15% 235

U and about 0.5% 234U, resulting in a dose estimate for 

one year of work of about 20 rad or 400 rem over 30 years (Polednak 1982). 

In light of the historical information cited in this review, the use an average air concentration of 

150 ug/m3

 by Polednak el al does not appear to be claimant friendly, and should be given 

particular attention by NIOSH.