December 8, 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 Rocky Flats Plant 

Dear Mr. Staudt: 

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

Rocky Flats Plant, consisting of our evaluation of “Revision 0” of six technical basis documents (TBDs) 

for introduction, 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. 

While this report was completed in October 2005, it was submitted to the Department of Energy on 

November 3, 2005, for purposes of confirming that there is no classified or Unclassified Controlled 

Nuclear Information (UCNI) national security information inadvertently included. DOE returned this 

draft report with such certification in a December 1, 2005 letter. 

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 (ABRWH) on one key finding contained in the 

report at the Advisory Board meeting in Knoxville, Tennessee, on October 17, 2005. This finding 

addressed our concern over NIOSH’s use of the median MDA (minimum detectable activity) values for 

plutonium and americium, which appear unduly low and likely to yield body burdens or organ doses that 

would be non-conservative, given the uncertainties involved. This finding, and its technical basis, was 

taken as an excerpt from the draft report and provided NIOSH for review prior to the November 16th issue 

resolution meeting held by the ABRWH working group in Cincinnati. 

In accordance with an informal agreement reached at the November 16th meeting, we are preparing an 

issue resolution matrix of the findings cited in this report and will provide it to the Advisory Board and 

NIOSH by December 14, 2005. An issue resolution meeting would presumably be scheduled at a 

mutually agreeable date and time thereafter. 
Mr. David Staudt 

December 8, 2005 

Page 2 

As indicated in earlier communications, we greatly appreciate the assistance and cooperation provided by 

Brant Ulsh and his team during the course of our review. 

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/001/08)
Draft 

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH 

National Institute of Occupational Safety and Health 

Rocky Flats Plant Site Profile Review 

Contract No. 200-2004-03805 

Task Order No. 1 

SCA-TR-TASK1-0008

Prepared by 

S. Cohen & Associates 

6858 Old Dominion Drive, Suite 301 

McLean, Virginia 22101 

Saliant, Inc.

5579 Catholic Church Road 

Jefferson, Maryland 21755

December 8, 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.

 
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Document No. 

SCA-TR-TASK1-0008 

Effective Date: 

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S. COHEN & ASSOCIATES:

Technical Support for the Advisory Board on 

Radiation & Worker Health Review of 

NIOSH Dose Reconstruction Program Revision No. 0 

ROCKY FLATS PLANT SITE PROFILE 

REVIEW Page 2 of 261 

Task Manager: 

________________________ Date: ___________ 

Joseph Fitzgerald 

Project Manager: 

________________________ Date: ___________ 

John Mauro 

Supersedes: 

N/A 

 
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TABLE OF CONTENTS 

Acronyms and Abbreviations ..........................................................................................................8 

1.0 Executive Summary...........................................................................................................11 

1.1 Summary of Strengths............................................................................................14 

1.2 Summary of Findings.............................................................................................15 

1.3 Observations ..........................................................................................................18 

1.4 Opportunities for Improvement .............................................................................19 

2.0 Scope and Introduction ......................................................................................................22 

2.1 Review Scope.........................................................................................................22 

2.2 Review Approach...................................................................................................24 

2.3 Report Organization...............................................................................................25 

3.0 Assessment Criteria and Methods......................................................................................26 

3.1 Objectives ..............................................................................................................26 

3.1.1 Objective 1: Completeness of Data Sources.............................................26 

3.1.2 Objective 2: Technical Accuracy..............................................................26 

3.1.3 Objective 3: Adequacy of Data.................................................................27 

3.1.4 Objective 4: Consistency Among Site Profiles.........................................27 

3.1.5 Objective 5: Regulatory Compliance........................................................27 

4.0 Site Profile Strengths.........................................................................................................32 

4.1 Completeness of Data ............................................................................................32 

4.2 Adequacy of Data ..................................................................................................32 

4.3 Technical Accuracy/Claimant Favorability ...........................................................33 

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

4.5 Regulatory Compliance .........................................................................................35 

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

5.1 Issue 1: Minimum Detectable Activity.................................................................36 

5.2 Issue 2: Occupational Internal Dose .....................................................................39 

5.2.1 Implications of High-Fired Plutonium in Bioassay Measurements for 

Acute and Chronic Intakes.........................................................................40 

5.2.2 Source Term...............................................................................................42 

5.2.3 Particle Size ...............................................................................................42 

5.2.4 Solubility....................................................................................................43 

5.2.5 Bioassay Data – Uncertainties and Minimum Detectable Activities.........50 

5.2.6 Lung Count Data........................................................................................53 

5.2.7 Other Bioassay Data ..................................................................................56 

5.2.8 Co-worker Dose Assignment.....................................................................59 

5.3 Issue 3: Potential for Ingestion Exposure Pathway...............................................59 

5.3.1 Plutonium Example....................................................................................59 

5.3.2 Americium Example ..................................................................................62 

5.3.3 Uranium Example ......................................................................................64 

5.4 Issue 4: Other Radionuclides................................................................................66 
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5.4.1 Inadequate Information Regarding Recycled Uranium.............................66 

5.4.2 Inadequate Information Regarding Highly-Enriched Uranium Storage 

Vulnerabilities............................................................................................67 

5.4.3 Inadequate Information Regarding the Processing of Uranium-233 .........68 

5.5 Issue 5: External Dose Calculation and Methodologies.......................................69 

5.5.1 Areas of Concern in Records and Dose Reconstruction............................69 

5.5.2 Job Exposure Matrix Study........................................................................71 

5.5.3 Below 10 mrem Dose Reported as Zero....................................................71 

5.5.4 Badges Only Calibrated for One Work Location ......................................71 

5.5.5 Exposure Geometry ...................................................................................71 

5.5.6 Angular Dependence for Beta/Gamma, NTA Film, and TLDs.................72 

5.5.7 Use of Dosimeters under Lead Aprons......................................................72 

5.5.8 Location of Dosimeters When Working Around Gloveports/boxes..........73 

5.5.9 Elevated Ambient Levels of External Radiation Affect Net Dose ............73 

5.6 Issue 6: Neutron Dosimetry and Exposures..........................................................73 

5.6.1 The Likelihood of Unmeasured Neutron Exposures .................................73 

5.6.2 Use of Neutron Track Plates and Uncertainties in Dosimetry, 1951–

1957............................................................................................................74 

5.6.3 NTA Neutron Energy Threshold May Lead to Missed Dose in RFP 

Records ......................................................................................................76 

5.6.4 NDRP Report Lacks Important Dose Reconstruction Information ...........77 

5.6.5 Neutron Dose Multiplication Factor..........................................................78 

5.7 Other External Dosimetry Issues...........................................................................78 

5.7.1 Unmonitored Individuals...........................................................................78 

5.7.2 Photon and Beta Dose Determination........................................................79 

5.7.3 Extremity Dose ..........................................................................................80 

5.7.4 Industrial X-ray Units and Neutron Generators.........................................81 

5.7.5 Decontamination and Decommissioning Activities Not Addressed..........82 

5.8 Issue 8: Historic Operations and Radiological Controls.......................................82 

5.8.1 Inadequate listing of Radionuclides of Concern........................................82 

5.8.2 The Need for Incident Database Access for Dose Reconstruction............83 

5.8.3 Inadequate Description of Post-Production Mission at Rocky Flats.........84 

5.8.4 Inadequate Description of Nuclear Plutonium Storage Activities and 

the Absence of Related Dose Reconstruction Guidance............................84 

5.8.5 The Need to Utilize Job Exposure Matrix Information .............................87 

5.9 Issue 9: Occupational Environmental Dose..........................................................88 

5.9.1 Soil Resuspension ......................................................................................88 

5.9.2 Other Radionuclides...................................................................................91 

5.9.3 Episodic Releases.......................................................................................91 

5.9.4 Data Adequacy and Completeness.............................................................92 

5.9.5 Timeline Needed for Phases of Operations, and Data Availability and 

Types..........................................................................................................94 

5.9.6 Particle Size ...............................................................................................95 

5.9.7 RATCHET Use is Suspect in the Vicinity of Closely Constructed 

Buildings....................................................................................................96 
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5.9.8 Need to Clarify if Maximum Estimates of Annual Intakes are Applied....96 

5.10 Issue 10: Occupational Medical Dose...................................................................97 

5.10.1 Guidelines Needed on What Constitutes Occupational Medical 

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

5.10.2 Potential for Other Types of X-ray Exposures ..........................................98 

5.10.3 Frequency and Types of X-ray Exposure is Derived from Other Sites.....99 

5.10.4 The Determination of Machine and Technician Uncertainties................100 

5.10.5 Use of Screens, Grids, and Impact of Off-Site Medical X-rays are Not 

Considered ...............................................................................................101 

5.11 Issue 11: Completeness and Adequacy of Data for Dose Reconstruction..........101 

5.11.1 General.....................................................................................................101 

5.11.2 Completeness of External Exposure Data................................................102 

5.11.3 Completeness of Internal Monitoring Data..............................................104 

5.11.4 Dosimetry Files........................................................................................106 

5.11.5 Data Integrity Concerns...........................................................................107 

5.11.6 Additional Issues Pertinent to External and Internal Dose 

Reconstruction .........................................................................................109 

6.0 Overall Adequacy of the site profile as a basis for dose reconstruction..........................116 

6.1 Satisfying the Five Objectives.............................................................................116 

6.1.1 Objective 1: Completeness of Data Sources...........................................116 

6.1.2 Objective 2: Technical Accuracy............................................................118 

6.1.3 Objective 3: Adequacy of Data...............................................................119 

6.1.4 Objective 4: Consistency Among Site Profiles.......................................120 

6.1.5 Objective 5: Regulatory Compliance......................................................121 

6.2 Usability of Site Profile for Intended Purpose.....................................................122 

6.2.1 Ambiguous Dose Reconstruction Direction ............................................122 

6.2.2 Inconsistencies and Editorial Errors in the Site Profiles..........................123 

6.3 Unresolved Policy or Generic Technical Issues ..................................................123 

7.0 References........................................................................................................................125 

Attachment 1: NIOSH Technical Documents Considered During the Review..........................132 

Attachment 2: Site Expert Interview Summary..........................................................................133 

Attachment 3: Key Questions for NIOSH/ORAU Regarding Site Profile Documents..............134 

Attachment 4: Summary of Conference Call on SC&A Questions Provided to NIOSH...........147 

Attachment 5: Consistency Between Site Profiles......................................................................220 

Attachment 6: Preliminary Talking Points for Conference Call – September 6–8, 2005...........232 

Attachment 7: Editorial Errors in the Technical Basis Documents............................................248 

Attachment 8: Records Requested from Rocky Flats.................................................................250 

Attachment 9: Evaluation of Inhalation Super S vs. Type S and Type M..................................252 

Attachment 10: Likelihood of Unmeasured Neutron Exposures................................................259 
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LIST OF TABLES

Table 1. Dose Coefficients for 1-Year Equivalent Dose Assuming a Single Intake of 1 

Bq of Plutonium-239 by Inhalation ............................................................................ 43 

Table 2. Comparison of Committed Equivalent Doses for Plutonium-239 Type M and 

Type S Compounds per Bq Plutonium-239 Present in the 24-Hour Working 

Day Urine Sample, Collected at the End of 1-Year Exposure ................................... 45 

Table 3. 1-Year Uranium-234 Committed Equivalent Doses per Bq Uranium-234 

Present in the 24-Hour Working Day Urine Sample, Collected at the End of 1-

Year Exposure ............................................................................................................ 48 

Table 4. Estimated Intake due to 20-Years Chronic Inhalation of Plutonium-239 

(Type S, AMAD 5μm) Based on the 1⁄2 Median MDA Value of Urinalysis 

(0.285 dpm)................................................................................................................. 51 

Table 5. Missed Doses for the Organs due to Inhalation of Plutonium-239, Based on the 

1⁄2 Median MDA Value of Urinalysis ......................................................................... 51 

Table 6. Intake Fraction in Urine and Feces after the Inhalation of 1 Bq of Plutonium239 Type S Compound (Bq/Bq intake) for 1 μm AMAD.......................................... 55 

Table 7. 1-Year Plutonium-239 Committed Equivalent Doses per Bq Plutonium-239 

Present in 24-Hour Working Day Urine Sample, Collected at the End of 1-Year 

Exposure ..................................................................................................................... 60

Table 8. 1-Year Americium-241 Committed Equivalent Doses per Bq Americium-241 

Present in 24-Hour Working Day Urine, Collected at the End of 1-Year 

Exposure ..................................................................................................................... 62

Table 9. 1-Year Uranium-234 Committed Equivalent Doses per Bq Uranium-234 

Present in 24-Hour Working Day Urine, Collected at the End of 1-Year 

Exposure ..................................................................................................................... 64

Table 10. Radionuclide contaminants in Depleted Uranium shipped to INEEL from

Rocky Flats and Fernald ............................................................................................. 67 

Table 11. Radioactive Materials at the Rocky Flats Plant.......................................................... 83 

 
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LIST OF FIGURES 

Figure 1. 1-Year Committed Equivalent Dose for All Organs due to 1-Year Chronic 

Inhalation of Plutonium-239 Type M and Type S Compounds ................................. 46 

Figure 2. 1-Year Committed Equivalent Dose for Systemic Organs due to 1-Year 

Chronic Inhalation of Plutonium-239 Type M and Type S Compounds.................... 46 

Figure 3. 1-Year Committed Equivalent Dose for All Organs due to 1-Year Chronic 

Inhalation of Uranium-234 Type S, Type M and Type S Compounds ...................... 49 

Figure 4. 1-Year Committed Equivalent Dose for Systemic Organs due to 1-Year 

Chronic Inhalation of Uranium-234 Type S, Type M and Type S Compounds......... 49 

Figure 5. Default Retention Groups (weak, moderate, strong and avid) for Radionuclides 

Deposited in Intramuscular Wound Sites................................................................... 58 

Figure 6. Schematic Compartmental Representation of the Draft NCRP Wound 

Biokinetic Model ........................................................................................................ 58 

Figure 7. 1-Year Committed Equivalent Dose for Organs due to 1-Year Chronic Intake of

Plutonium-239, in Sv per Bq Excreted in 24-Hour Working Day Urine Samples 

after 1 Year of Work................................................................................................... 61 

Figure 8. 1-Year Committed Equivalent Dose for Systemic Organs due to 1-Year 

Chronic Intake of Plutonium-239, in Sv per Bq Excreted in 24-Hour Working 

Day Urine Samples after 1 Year of Work .................................................................. 61 

Figure 9. 1-Year Committed Equivalent Dose for Organs due to 1-Year Chronic Intake of

Americium-241, in Sv per Bq Excreted in 24-Hour Working Day Urine 

Samples after 1 Year of Work .................................................................................... 63 

Figure 10. 1-Year Committed Equivalent Dose for Systemic Organs due to 1-Year 

Chronic Intake of Americium-241, in Sv per Bq Excreted in 24-Hour Working 

Day Urine Samples after 1 Year of Work .................................................................. 63 

Figure 11. 1-Year Committed Equivalent Dose for Organs due to 1-Year Chronic Intake of

Uranium-234, in Sv per Bq Excreted in 24-Hour Working Day Urine Samples 

after 1 Year of Work................................................................................................... 65 

Figure 12. 1-Year Committed Equivalent Dose for Systemic Organs due to 1-Year 

Chronic Intake of Uranium-234, in Sv per Bq Excreted in 24-Hour Working 

Day Urine Samples after 1 Year of Work .................................................................. 65 
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ACRONYMS AND ABBREVIATIONS 

Advisory Board Advisory Board on Radiation and Worker Health 

A-P Anterior-Posterior 

AEC Atomic Energy Commission 

AED Activity Equivalent Diameter 

AMAD Activity Median Aerodynamic Diameter 

Bq Becquerel 

BR Brass 

CATI Computer-Assisted Telephone Interview 

CD Cadmium

CDC Centers for Disease Control and Prevention 

CDPHE Colorado Department of Public Health and Environment 

CF Calibration Factors 

CFR Code of Federal Regulations

Ci Curie 

CWA Clean Water Act 

D&D Decontamination and Decommissioning 

DNFSB Defense Nuclear Facilities Safety Board 

DOE Department of Energy 

dpm Disintegrations per Minute 

DR Dose Reconstruction 

DTPA Diethylenetriaminepentaacetate 

DU Depleted Uranium

EALER Elevated Ambient Levels of External Radiation 

EDTA Ethylenediaminetetraacetic acid 

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000 

EPA Environmental Protection Agency 

ESE Entry Skin Exposure 

EU Enriched Uranium

FBI Federal Bureau of Investigation 

GeLi Germanium Lithium

 
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GAO General Accounting Office 

HCEC House Committee on Energy and Commerce 

HEU Highly-Enriched Uranium 

HHS Health and Human Services 

IAAP Iowa Army Ammunition Plant 

ICRP International Commission on Radiological Protection 

IMBA Integrated Modules for Bioassay Analysis 

INL Idaho National Laboratory 

IREP Interactive RadioEpidemiological Program

keV Kiloelectron Volt 

LAP Laboratory Accreditation Program

LOD Limit of Detection 

μCi Microcurie 

MCW Mallinckrodt Chemical Works 

MDA Minimum Detectable Activity 

MMD Mass Median Diameter 

mR Milliroentgen 

mrad Millirad 

mrem Millirem

NCRP National Council on Radiation Protection and Measurements 

NDRP Neutron Dose Reconstruction Protocol 

NDT Nondestructive Testing 

NIOSH National Institute for Occupational Safety and Health 

NOCTS NIOSH Occupational Claims Tracking System

NPDES National Pollutant Discharges Elimination System

NTA Eastman Kodak Nuclear Track Film Type A 

NTS Nevada Test Site 

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 
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OW Open Window 

PA Posterior-Anterior 

PFG Photofluorography 

PIC Pocket Ionization Chamber 

POC Probability of Causation 

POS Process Operation Sheets 

PPE Personnel Protective Equipment 

ppm Parts Per Million 

RAC Risk Assessment Corporation 

RadCon Radiological Control 

RCRA Resource Conservation and Recovery Act 

RDGs Radiation Generating Devices 

RFP Rocky Flats Plant 

RU Recycled Uranium

SAAM Selective Alpha Air Monitor 

SC&A S. Cohen and Associates 

SRS Savannah River Site 

STP Sewage Treatment Plant 

TBD Technical Basis Document 

TIB Technical Information Bulletin 

TLD Thermoluminescent Dosimeter

TLND Thermoluminescent Neutron Dose 

TRU Transuranics 

Y-12 Plant Y-12 National Security Complex 

ZPPR Zero Power Plutonium Reactor
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1.0 EXECUTIVE SUMMARY 

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

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

Flats Plant (RFP) developed by the National Institute for Occupational Safety and Health 

(NIOSH). This audit was conducted during the period July 15 through September 30, 2005, in 

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

statutory responsibility under the Energy Employees Occupational Illness Compensation 

Program Act of 2000 (EEOICPA) to conduct such reviews and advise the Secretary of Health 

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

Located in Golden, Colorado (sixteen miles northwest of Denver), the RFP covered 6,550 acres 

of land, and included more than a hundred buildings and other structures. The site is composed 

of a 384-acre industrial area surrounded by an operational buffer zone. Rocky Flats was built in 

1951 and began operations in 1952. The plant’s primary mission was to produce plutonium

weapon components for nuclear weapons. Production activities involved the fabrication of 

plutonium, uranium, beryllium, and stainless steel parts. Other activities included chemical 

processing for recovery of plutonium from scrap material; research and development work in 

metallurgy; machining; assembly; nondestructive testing; coatings; remote engineering; and 

chemistry and physics. Parts were made at the facility and shipped elsewhere for assembly 

(Norton et al. 1992). Dow Chemical Company (1952–1975), Rockwell International 

Corporation (1975–1989), EG&G Corporation (1990–1995), and Kaiser-Hill (1995–present) 

have operated the plant for the Department of Energy (DOE). The plant ceased plutonium

operations in November 1989, and began its transformation into an environmental restoration 

cleanup site in 1993. The plutonium production facilities have been decontaminated and 

dismantled, with overall site closure scheduled for 2006. 

This review included unclassified interviews of site experts by SC&A personnel at the Denver 

Federal Center,1

 as well as conference calls between Oak Ridge Associated Universities (ORAU) 

and SC&A counterparts regarding specific TBDs that make up the RFP 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. 

In addition to this TBD, other RFP and DOE Complex-wide TBDs were reviewed by SC&A. A 

complete list of the RFP TBDs, as well as supporting Technical Information Bulletins (TIBs), 

that were reviewed by SC&A is provided in Attachment 1. 

This review found the NIOSH site profile (and its constituent TBDs) for RFP to be an adequate 

accounting for most, but not all, of the plutonium exposure and dosimetric history of the plant, 

falling short in obtaining some key data and fully characterizing several underlying issues 

fundamental to guiding dose reconstruction. It is clear that NIOSH recognizes some of these 

1 While the review was unclassified, all material generated from worker interviews and document review 

were submitted to classification screening to assure that no sensitive information was inadvertently included in this 

report. 
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issues and has indicated that a number of ongoing initiatives are underway to obtain missing data 

or generate needed guidance (e.g., new TIB addressing highly insoluble oxides and guidance on 

co-worker dose assignment assumptions and methodology). It is also clear that some additional 

implementation guidance may reside in “workbooks,” which are developed by NIOSH as a 

means to provide implementation guidance to the dose reconstructor; these resources were not 

within the stipulated scope of this review for RFP, but will be reviewed as part of the FY2006 

program. 

One of SC&A’s key concerns is NIOSH’s use of median MDA (minimum detectable activity) 

values for plutonium and americium, which appear unduly low and likely to yield body burdens 

or organ doses that would be non-conservative, given the uncertainties involved. Given the 

limited data and uncertainties of the key variables (e.g., sample count time, detector counting 

efficiency, self-absorption, and various sampling assumptions) from which MDAs are defined, 

SC&A understands the need to apply certain assumptions to bridge these gaps in information. 

However, the concern is that a number of these assumptions are not adequately supported and 

may not be claimant favorable. Likewise, the use of these assumed median MDA values, 

themselves, in dose reconstruction, may be inappropriate. This is because urine activity levels 

monitored at RFP for plutonium and americium were likely well in excess of assumed median 

MDA values, most notably where workers were assigned “zero” or “background” readings in the 

past when urinalysis results were found to be less than 10% of the tolerance level. SC&A 

believes more conservative assumptions should be applied in the formulation of MDA values 

that would more realistically reflect the range of uncertainties involved. 

Likewise, a number of historical issues and discrepancies cast doubt on the validity, and in some

instances, the integrity, of dose records being relied upon for EEIOCPA dose reconstruction―

ones which are not adequately reflected in the TBDs that make up the site profile. Evidence 

exists that there is a potential for missed occupational external, internal, and environmental dose, 

due to the incompleteness and questionable quality of data being used in dose reconstruction. 

These missed doses may have occurred as a result of incomplete monitoring records and 

inadequate monitoring techniques. There may also be a data integrity issue associated with the 

external dose record. 

The SC&A review finds that the internal dosimetry TBD (Falk 2004) lacks definitive direction in 

some instances and has gaps that need to be addressed. There is limited guidance for use by the 

dose reconstructor regarding the process and assumptions that should be used to calculate 

internal dose. Notably, this TBD does not provide clear guidance for assessment of dose for 

unmonitored workers, nor does it specifically address what “missed dose” may exist and how it 

is to be addressed. The use of the assumed default particle size of 5 μm AMAD needs to be 

reconsidered for those RFP operations for which actual particle size measurements exist (e.g., an 

0.3 μm mass median diameter for airborne particles involved in at least two fires at RFP, which 

may be typical of “high-fired” plutonium generated in processes involving temperatures 

exceeding 400°–600° C). The approaches regarding solubility need to be reviewed, particularly 

for Type “S” or “Super-S” plutonium compounds whose high insolubility may lead to more 

exposure to gastrointestinal and respiratory tract organs. The assumptions (e.g., full equilibrium) 

regarding the methodologies to assess the internal exposure to depleted uranium based on 
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estimates of 238U activity may not be claimant favorable for some circumstances. The internal

dose potential for the ingestion pathway was not considered. 

The external dosimetry sections of the TBDs and supporting TIBs, while providing pertinent 

historical data and technology information on dosimetry systems and records, fall short in 

addressing potential missed and unmonitored dose, particularly in the early years of operation 

(1950s through mid-1960s). The use of neutron track plates, and the uncertainties in neutron 

track counting with these and NTA film, indicates there may be important missed dose that was 

not entered into claimant records. The Neutron Dosimetry Reconstruction Project Protocol 

(NDRP) was designed to remedy these deficiencies by recounting original tracks and providing a 

corrected estimated individual neutron dose. However, the recently issued NDRP report does 

not cover non-Pu workers, nor is it applicable to unmonitored or non-neutron workers. There is 

a resultant need to use neutron/photon ratios and film/TLD comparisons to correctly determine 

past neutron doses. The TBD (Langsted 2004) only briefly addressed these two issues, which 

could be important in generating correction factors for under-monitored workers or for 

monitored worker missed dose. 

The issue of unmonitored workers, particularly from the early 1950s to the early 1960s, is not 

developed in the TBDs from the standpoint of the extent of the problem and how dose estimation 

would be handled, particularly the application of co-worker doses. Assignment of dosimeters to 

workers using the “10% of the radiation protection guide” policy was apparently based on a 

priori administrative decisions, not necessarily on actual survey results, as evidenced by gaps 

identified in the NDRP report and supported by site expert interviews. This underscores the need 

for NIOSH to corroborate, for the purpose of co-worker dose assignment, the extent to which 

badged workers represented the maximally exposed individuals during the early years of plant 

operation. 

The RFP occupational medical dose TBD does not adequately address the contribution of 

historic radiation exposure from occupationally necessitated medical x-ray exposure of workers 

at Rocky Flats. By confining consideration to one pre-employment and a chest x-ray annually, it 

may underestimate potential dose by excluding consideration of x-ray procedures required for 

special requirements, e.g., respiratory protection certification, and work related injuries. The 

TBD does not document that x-ray equipment did not exceed the diagnostic range, or whether it 

was greater than 80 kVp, as no evidence of radiation surveys exist prior to 2001. The TBD does 

not adequately address the potential dose from photofluorography (PFG) units, since some

workers may have had multiple PFG exams in a year. 

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

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

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

review criteria. Several of the issues were designated as 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/ORAU has supplemented the RFP site profile documents with a number of TIBs that 

provide further direction to the dose reconstructor. Attachment 1 lists four TIBs that all have 

direct bearing on the RFP 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 RFP TBDs to the dose 

reconstruction process. NIOSH is working on an additional TIB that will specifically address 

uptakes of Super Type S.

The TBD traced the history of the various functions and processes at the RFP that could 

contribute to worker radiation dose, as well as describing the major use of the many individual 

buildings at the complex. In addition, operation accidents and incidents through November 1976 

are listed in Attachment 2C of the site description TBD. The two major fires occurring in 1957 

and 1969 were covered in some detail. Inclusion of major accidents and incidents is an 

improvement over previous site profiles. Overall, with the notable exception of recycled 

uranium and other radiological materials shipped onsite, the site description TBD provides an 

accurate and reasonable characterization of the history and operational activities at the RFP for 

the period of production. 

NIOSH/ORAU provided a detailed overview of routine and neutron dosimeters from the 

inception of production through the present. The external TBD includes a discussion of the 

effectiveness of dosimeters relative to low-energy photon and neutron spectra in plutonium

production areas. Techniques are included to assist dose reconstuctors in determining radiation 

type and energy. Methodologies for missed external dose calculations are covered for both 

unmonitored individuals and individuals with zero dose values. Site-specific correction factors 

have been developed to compensate for dosimeter inadequacy (e.g., NTA energy dependence) 

and uncertainty. 

NIOSH/ORAU has provided a detailed history of the in-vivo and in-vitro programs at Rocky 

Flats. Techniques for plutonium, americium, uranium, and gross alpha bioassay, as well as 

limited data on other radionuclides, have been included in the internal TBD. The TBD has 

provided information on solubility, particle size, and isotopic composition for major source 

terms. Detailed descriptions of the lung counting systems have been provided for reference. 

NIOSH/ORAU has provided helpful information on the interpretation of records, as well as

examples of these records. This is perhaps the most helpful portion of the internal and external 
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TBDs to dose reconstruction. NIOSH/ORAU is encouraged to continue including information 

on interpretation of records and provision of example records in this and other site profiles. 

1.2 SUMMARY OF FINDINGS 

Finding 1. Suggested use of urine bioassay MDA values for plutonium and americium 

appear low and are likely to yield body burdens/organ doses that are proportionately low

and, therefore, claimant unfavorable. NIOSH’s use of median MDA (minimum detectable 

activity) values for plutonium and americium appear unduly low and likely to yield body burdens 

or organ doses that would be non-conservative, given the uncertainties involved. Given the 

limited data and uncertainties of the key variables (e.g., sample count time, detector counting 

efficiency, self-absorption, and various sampling assumptions) from which MDAs are defined, 

SC&A understands the need to apply certain assumptions to bridge these gaps in information. 

However, the concern is that a number of these assumptions are not adequately supported and 

may not be claimant favorable. Likewise, the use of these assumed median MDA values, 

themselves, in dose reconstruction, may be inappropriate. This is because urine activity levels 

monitored at RFP for plutonium and americium were likely well in excess of assumed median 

MDA values, most notably where workers were assigned “zero” or “background” readings in the 

past when urinalysis results were found to be less than 10% of the tolerance level. SC&A 

believes more conservative assumptions should be applied in the formulation of MDA values 

that would more realistically reflect the range of uncertainties involved. 

Finding 2: Internal dosimetry TBD lacks definitive direction in some instances and has 

gaps that need to be addressed. There is limited guidance for use by the dose reconstructor 

regarding the process and assumptions that should be used to calculate internal dose. Notably, 

this TBD does not provide clear guidance for assessment of dose for unmonitored workers, nor 

does it specifically address what “missed dose” may exist and how it is to be addressed. The use 

of the assumed default particle size of 5 μm AMAD needs to be reconsidered for those RFP 

operations for which actual particle size measurements exist (e.g., a 0.3 μm mass median

diameter for airborne particles involved in at least two fires at RFP). The approaches regarding 

solubility need to be reviewed, particularly for Type “S” or “Super-S” plutonium compounds 

whose high insolubility may lead to more exposure to gastrointestinal and respiratory tract 

organs. Uncertainties are not addressed in the TBD regarding the 241

Am assay of plutonium

processed at RFP and how lung counting was calibrated to these values. The assumptions (full 

equilibrium) regarding the methodologies to assess the internal exposure to depleted uranium

based on estimates of 238U activity may not be claimant favorable for some circumstances. The 

sensitivity of the bioassay methods was not adequate to detect incidental intakes of insoluble 

compounds.

Finding 3: Interpretation of NTA film data for workers who were not included in NDRP 

re-evaluation is not evident; guidance on use of neutron/photon ratios not available. The 

use of neutron track plates, and the uncertainties in neutron track counting with these and NTA 

film, indicates there may be important missed dose that was not entered into claimant records. 

Although the TBD recommends the use of the Neutron Dosimetry Reconstruction Project 

(NDRP) assessed doses, this data has only recently become available for use in dose 
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reconstruction. Further discussion of this data and how it is to be used should be included in 

future revisions of the TBD. The NDRP report does not cover non-plutonium worker, nonneutron workers, or workers not monitored with NTA film. There is a resultant need to use 

neutron/photon rations and/or film/TLD comparisons to correctly determine past neutron doses. 

The TBD (Langsted 2004) only briefly addressed these two issues, which could be important in 

generating correction factors for under-monitored workers or for monitored-worker missed dose. 

This stems from a concern that workers exposed to neutrons in the early days of NTA film use at 

lower energy levels (between 0.25 and 0.4 MeV) will have missed dose that has not been 

accounted for in the TBD. The NTA film could not detect neutrons below a 0.8 MeV to 1.0 

MeV threshold. NIOSH has attempted to obtain data from the Job Matrix Study that was 

developed by the University of Colorado (Ruttenber 2003). This information has not been made 

available and NIOSH is taking additional actions to gain access. 

Finding 4: Unclear treatment in TBD of personal dosimeter placement and angular 

dependence. In the dose reconstruction process, the assignment of isotropic (ISO) or rotational 

(ROT) instead of anterior-posterior (AP) geometry in the TBD may not reflect the true radiation 

dose to some workers. In addition, the issue of angular dependence for different types of 

radiation and different dosimetry systems in use through the years is not sufficiently addressed in 

the TBD. There was a potential for partial body exposures in excess of the whole-body 

dosimeter reading (e.g., exposure to the head, face, and uncovered body parts during use of lead 

aprons; wearing dosimeters at chest height for glovebox operations). This issue has not been 

identified in the TBD. There is evidence that “elevated ambient levels of external radiation” 

(EALER) occurred at RFP with routine day-to-day storage of dosimeters, an issue of which 

NIOSH is aware, but is not addressed adequately in the TBD. 

Finding 5: The RFP occupational medical dose TBD does not adequately address the 

contribution of historic radiation exposure from occupationally necessitated medical x-ray 

exposure of workers at Rocky Flats. The TBD includes only pre-employment and annual 

routine chest exams as representing the total occupational medical dose. Using this 

methodology, special job-required x-rays, such as x-rays for respiratory protection certification, 

and special exams for asbestosis and beryllium workers are potentially missed. Also, ORAUTOTIB-0006 (Kathren and Shockley 2005), which is the basis document for all the TBD 

occupational medical exposures, further suggests the need for inclusion of special exams and 

termination exams, which are not included in the RFP TBD. Also omitted are x-ray exposures 

resulting from work-related injuries. A review of RFP medical records shows that there is no 

formal documentation to prove that workers did not get multiple x-ray exposures for work 

related to asbestos and beryllium exposure. The TBD does not document that x-ray equipment 

did not exceed the diagnostic range, or whether it was greater than 80 kVp, as no evidence of 

radiation surveys exist prior to 2001. The TBD does not adequately address the potential dose 

from photofluorography (PFG) units, since some workers may have had multiple PFG exams in 

a year. Overall, the TBD relies too heavily on the general information contained in ORAUOTIB-0006 (Kathren and Shockley 2005), because there is no documentation on x-ray protocols 

or machine calibration requirements prior to 2001. 

 
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Finding 6: The site profile, while incorporating methodologies for assignment of missed 

dose, has not adequately bound exposure conditions, compensated for calibration errors 

and technology deficiencies, and addressed possible data integrity issues, all of which may 

contribute to missed dose. A number of historical issues cast doubt on the validity of recorded 

doses. In the case of external dosimetry, these issues included problems with algorithms, 

calibrations, placement of dosimeters in relation to aprons, individuals not wearing and/or 

improperly wearing their dosimeters, and uncertainty with the conditions under which zero or 

null doses were assigned. In some cases zeros may have been recorded in lieu of legitimate dose 

estimates. Neutron exposures to workers during peak production in the 1950s and 1960s were 

likely to be unmeasured – particularly given; (a) relatively primitive neutron counting

techniques; (b) the informal, non-binding nature of criticality safety and neutron exposure 

requirements, and (c) the small staff assigned at the time to address these hazards. With internal 

dosimetry, there were individuals not monitored and delinquencies in submittal of bioassay 

samples. The source data for calculation of airborne environmental dose is questionable based 

on effectiveness of environmental air sampling. The limited information on job titles, skills and 

tasks, work locations, time spent in the area, and radionuclides of concern further complicate 

issues, making it difficult to differentiate between radiological and non-radiological workers. 

Finally, the concentration of 241Pu and the age of plutonium, itself, from recycled warheads need 

to be studied to ascertain the validity of plutonium intake estimates used in the internal dose 

TBD. 

Finding 7: The internal TBD does not consider potential contribution of ingestion 

pathway. The internal dose potential for ingesting radionuclides has not been considered in the 

occupational internal dosimetry TBD (Falk 2004). The ingestion pathway should not be ignored, 

except for the organs related to the respiratory tract where the dose from inhalation 

predominates. For other organs, ingestion dose is often higher than any dose received from

inhalation. This is particularly true for plutonium, americium, and uranium. The ingestion 

pathway was also not considered in the derivation of co-worker dose data. The use of bioassay 

results to back-calculate intake and doses will produce higher internal exposures for certain 

organs if the ingestion pathway is taken into account. 

Finding 8: TBDs do not adequately address potential exposure contribution of recycled 

uranium and other radiation sources shipped onsite. The RFP site description TBD (Flack 

and Meyer 2004) does not provide an accurate assessment of the potential risks associated with 

recycled uranium. According to a U.S. Department of Energy report (DOE 2000), the DOE’s 

Idaho National Environmental and Engineering Laboratory (INEEL) shipped quantities of 236U 

recovered from previously irradiated reactor fuel in 1955 to the RFP. This represents a potential 

for significant gamma fields and a potential source of missed dose for RFP workers. Also the 

TBD makes only a passing reference to the processing of 233U, but it is known that Rocky Flats 

received 233

U uranyl nitrate from Oak Ridge for processing into kilogram quantities of metal 

shapes from the mid 1960s to the early 1980s. Of particular concern is exposure to 232U 

contaminants that are co-produced with 233

U irradiation of thorium and are substantially more 

radioactive than 238U. The potential dose attributable to neptunium, thorium, curium, and tritium

has not been addressed, and more specific guidance to dose reconstructors is needed, given the 

absence of bioassay data or air concentration data. 
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Finding 9: The occupational environmental TBD does not adequately address potential 

environmental exposure from ambient airborne releases and resuspension of contaminated 

soil. Routine and episodic ambient airborne releases have been brought into question, based on 

the adequacy of air monitoring results. Incidental releases determined by the state of Colorado 

are higher than the values used for the 1957 and 1969 fires in the TBD, resulting in nonclaimant-favorable assumptions. The dose from resuspension of soil contaminated with 

plutonium, americium, and other radionuclides (e.g., 238Pu, 137

Cs, and 237Np) needs to be taken 

into consideration for soil contamination areas throughout the site, and should not be limited to 

the 903 Pad without some justification why other inactive waste sites (108 in all) are not 

included. SC&A also believes that resuspension of 239+240Pu and 241Am throughout the site could 

be an important contributor to ambient dose for both monitored and unmonitored dose. In 

particular, the TBDs do not clearly address how internal dose assessments will consider the 

contribution of resuspended plutonium and americium to worker dose. 

Finding 10: Hand and wrist doses are not adequately addressed in the external dosimetry 

TBD. The TBD needs to provide more information on how to account for extremity dose, since 

there was considerable hands-on work at RFP. Also, valid hand-to-wrist ratios are needed and 

could be useful in potentially much higher dose to workers’ hands. It is not apparent that the 

VARSKIN software includes the predominant beta emitter, 234mPa that is present at the RFP. 

Finding 11: The TBDs (Langsted 2004; Little and Meyer 2004) do not address the 

potentially significant doses from industrial x-ray and neutron generators for R&D and 

non-destructive (NDT) work done at RFP. These may represent potentially high radiation 

exposures sources with energy spectra for which a mismatch may have existed with dosimeter 

systems in use at the time calibrated for plutonium and uranium isotopes. Information, 

including the number of units, energies, periods of operations, and operating procedures, need to 

be identified, with some characterization performed of the potential radiation exposures 

involved, and if radiation doses were under-recorded or missed. 

1.3 OBSERVATIONS 

Observation 1: The RFP site profile does not address post-production (post-1992) 

decontamination and decommissioning activities and worker exposures. This period involved 

nuclear material storage, nuclear material stabilization and packaging, waste management, and 

decontamination and decommissioning, for which records show a history of contamination 

incidents and personnel exposure. Similarly, the environmental, safety, and health 

vulnerabilities associated with the storage of plutonium and high enriched uranium at RFP 

during these and previous time periods are not addressed. For example, in the 1990s, RFP held 

over metric tons of highly enriched uranium (HEU) consisting mostly of metals in the form of

pits, part samples, and scrap. 

Observation 2: The overlap in definition of phases of operation requires further study to 

identify dose from radionuclides such as tritium, thorium, enriched and depleted uranium

239+240

Pu, 241Pu and 241Am which can be related to significant releases. A timeline is needed to 
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distinctly delineate phases of operation, data types and availabilities, as well as data sources 

used. 

Observation 3: The use of the RATCHET air dispersion model is not able to take into account 

unexpected air flows around close-in buildings, where the height of those buildings may perturb 

estimates because of wake effects. This may be a confounding problem at RFP and needs to be 

addressed. 

Observation 4: With an appropriate wound dose model not available, the cited approach for 

estimating wound-related uptakes in the internal dose TBD is claimant favorable for relevant 

types of cancer, except for lymph nodes and skin cancers. A more claimant-favorable approach 

for these affected organs needs to be addressed, for which a recently proposed model (Guilmette 

and Durbin 2003) for wound-site retention of soluble radionuclides may be relevant. 

1.4 OPPORTUNITIES FOR IMPROVEMENT 

Incomplete Personal Monitoring Data: NIOSH should make further efforts to identify the 

dosimetry systems and recording methods used to record each worker’s exposure through the 

years, as these issues affect dose reconstruction. It is important that NIOSH succeed in gaining 

access to the University of Colorado’s Job Exposure Matrix to more effectively identify worker 

job categories and enhance the ability to reconstruct dose for all special work assignments at the 

RFP. These data provide important insights into exposure durations, practices in which 

exposures may have occurred, changes in job classifications over time, and job responsibilities. 

This study should also be used to assist in identifying workers who may have been exposed to 

radiation from work-related tasks and were not badged, especially during the early years. 

However, the data from this matrix are not a panacea for the need to identify data quality and 

gaps, and to recover available personnel dose and plant operational records as they still exist. 

NIOSH likewise needs to work on more detailed extremity dose assignment factors, such as 

whole body-to-wrist-to-hand conversion factors for both beta/gamma and neutron exposures. 

Internal dosimetry: In its ongoing revision of this TBD, NIOSH needs to consider 

strengthening it in terms of how it treats the various uncertainties, detection limits, and

limitations that have been identified, particularly for chronic low-level intakes, and for the 

broader range of radionuclides historically present in RFP operations. Coupled with supporting 

TIBs and other guidance documents, the TBD needs to provide clear guidance regarding 

pertinent assumptions and approaches for estimating dose for unmonitored workers, workers 

exposed to various plutonium compounds of varying solubility and particle size, and the 

ingestion pathway. 

Other Radionuclides: The TBD Site Description provides useful information regarding 

radionuclides handled in different buildings, but does not list the general types and quantities of 

radionuclides to which workers may have been exposed. Data compiled for offsite dose 

reconstruction for the surrounding public indicate that Rocky Flats handled at least 85 

radioisotopes. A more complete characterization of radionuclides and their quantities needs to 

be completed. The dose contribution from 233U and daughters should be evaluated and NIOSH 
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should place more emphasis and study on evaluating dose potential from radionuclides, such as 

tritium, thorium, enriched and depleted uranium 239+240Pu, 241Pu, and 241Am, which can be related 

to significant releases. 

Neutron Monitoring: NIOSH should consult the NDRP for more defensible neutron dose data 

and incorporate this into their dose reconstructions. For those not in the NDRP cohort, neutron 

exposure should be investigated, and a methodology developed to determine a claimantfavorable neutron dose. Additionally, more use of neutron/photon ratios and/or film/TLD 

comparisons could be made to arrive at more individualized neutron doses and more robust 

neutron dose multiplication factors for assigning doses to unbadged workers, rather than relying 

on assigning average doses derived from the general working population. 

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 

provides original perspectives and information concerning site practices and exposure histories. 

NIOSH has incorporated only a limited amount of worker input into the latest versions of the 

TBD. 

Occupational Medical Exposure: NIOSH should reconsider its interpretation of occupational 

medical exposure. The definition should be expanded to include all forms of medical exposures 

associated with job performance, to include illness and injury exams, return to work exams, 

special or certification exams, and work termination exams. NIOSH needs to further evaluate all 

medical and radiation survey records to better assess machine use and output parameters. 

Post-Production Mission at Rocky Flats: NIOSH should address dose reconstruction for 

workers involved in the post-production mission at Rocky Flats. The post-production mission, 

now in its 16th year, has involved nuclear material storage, nuclear material stabilization and 

packaging, environmental restoration, waste management, and decontamination and 

decommissioning. All of these activities have resulted in employees working in various 

radiation environments, which should be described and characterized relative to radiation 

exposures and dose reconstruction. 

Utilization of Incident Data for Dose Reconstruction: NIOSH should not rely on the a priori

assumption that the files of individual claimants contain all relevant information on incidents 

involving the claimant. For monitored workers, ORAU and NIOSH rely primarily on the worker 

dose record and the Computer Assisted Telephone Interview (CATI) for potential exposure data 

on radiological incidents. Workers have often complained of poor record keeping and 

fabrication of records, which have been confirmed in congressional and Defense Nuclear 

Facilities Safety Board (DNFSB) investigations. Although the Rocky Flats Site Profile provides 

a list of accidents and incidents, as it now stands, it does not have a dedicated database or 

assessment for off-normal exposures and incidents (e.g., uranium and plutonium fires, 

explosions, spills, etc.). In fact, the list provided in the TBD does not contain incidents past 

November 1976. Careful evaluation of incident record keeping practices may indicate problems 

with the integrity of individual worker data in some settings and periods. Available incident 

databases, even if potentially incomplete, should be incorporated into dose reconstruction efforts. 
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This is particularly important for workers exposed during and after hundreds of incidents, 

including fires, explosions, leaks and spills. The Former Radiation Worker Medical Surveillance 

Program at Rocky Flats, which includes 875 participants, indicates that some workers involved 

in incidents received lung committed dose equivalents from 600 rem to 12,000 rem and bone 

surface committed dose equivalents from 500 rem to 32,000 rem. However, this program is 

limited by its voluntary nature and does not necessarily capture the degree and extent of 

exposures from all relevant incidents. 
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2.0 SCOPE AND INTRODUCTION 

The review of the Rocky Flats Plant (RFP) site profile was conducted from July 15 through 

September 30, 2005, in support of the Advisory Board on Radiation and Worker Health 

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

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

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

adequacy” of the EEOICPA program. This review included unclassified interviews of site 

experts by S. Cohen & Associates (SC&A) personnel at the Denver Federal Center, 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 RFP site profile. While 

the review was unclassified, 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. The Rocky Flats site officially ceased plutonium operations

in 1993 and was completing final closure activities at the time of this review. Therefore, no site 

activities were observable, and relevant records had been largely archived and not readily 

available during the site visit. There is currently an outstanding request for records from the RFP 

representatives. 

2.1 REVIEW SCOPE 

Under the 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 is mandated 

to conduct an independent review of the methods and procedures used by the National Institute 

for Occupational Safety and Health (NIOSH) and its contractors for dose reconstruction. As a 

contractor to the Advisory Board, 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 RFP site: 

• ORAUT-TKBS-0011-1, Technical Basis Document for the Rocky Flats Plant – 

Introduction, Rev. 00 (Little and Meyer 2004) 

• ORAUT-TKBS-0011-2, Technical Basis Document for the Rocky Flats Plant – Site 

Description, Rev. 00 (Flack and Meyer 2004) 

• ORAUT-TKBS-0011-3, Technical Basis Document for the Rocky Flats Plant – 

Occupational Medical Dose, Rev. 00 (Furman and Lopez 2004) 

• ORAUT-TKBS-0011-4, Technical Basis Document for the Rocky Flats Plant – 

Occupational Environmental Dose, Rev. 00 (McDowell-Boyer and Little 2004) 
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• ORAUT-TKBS-0011-5, Technical Basis Document for the Rocky Flats Plant – 

Occupational Internal Dose, Rev 00 (Falk 2004) 

• ORAUT-TKBS-0011-6, Technical Basis Document for the Rocky Flats Plant – 

Occupational External Dosimetry, Rev. 00 (Langsted 2004) 

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

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

Attachment 1. 

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

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

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

SC&A team did not include review of workbooks related to the RFP site profile. The Advisory 

Board did direct SC&A, beginning in FY2006, to conduct an evaluation of 20 such workbooks in 

concert with its site profile reviews under Task 1. The workbook(s) associated with RFP will be 

evaluated in that context following issuance of this review. 

Beyond the conduct of its independent interviews of site experts, the SC&A team is aware of and 

has requested access to a NIOSH database named “Top Hat,” which contains NIOSH/ORAUconducted worker interviews. It was the team’s understanding that the database is undergoing 

modification in order to make it available online, and that it would be available for access by 

November 2005. A formal request has been made for such access, and an addendum to this 

report will be provided based on the results of an evaluation of information from that source. 

SC&A, in support of the Advisory Board, has critically evaluated the RFP Site 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 SC&A Standard Operating Procedure for Performing 

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

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

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

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

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

reconstructions, bridging uncertainties, or correcting technical inaccuracies. This review does 
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not explicitly address the issue of radiation exposures to cleanup workers and decommissioning 

workers, as that is not addressed in the TBDs. 

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

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

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

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

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

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

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

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

the TBDs would provide for robust compensation decisions. 

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

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

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

methodologies is dosimeter readings and bioassay data, co-worker data and workplace 

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

2.2 REVIEW APPROACH 

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

completeness of data collected by NIOSH, the adequacy of existing RFP personnel and 

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

expert interviews were conducted as “non-classified” interviews in a secure facility at the Denver 

Federal Center and all notes taken were screened by RFP derivative classifiers. 

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

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

Attachment 2 provides summaries of the interviews conducted by SC&A by teleconference or in 

person in Denver during the course of this review. The site experts included current and former 

staff from radiation control, operations, environmental monitoring, maintenance, and other 

organizations, as well as other site experts knowledgeable of particular elements of the RFP 

environmental safety and health program. These interviews were conducted during the course of 

the RFP site profile review. These individuals were given the opportunity to review the 

interview summary for accuracy. This is an important safeguard against missing key issues or

misinterpreting some vital piece of information. A master summary of all interviews conducted 

was compiled and is included in Attachment 2. The summary is not intended to be a verbatim

transcript, but paraphrases conversations held with site experts. References to specific site 

experts have been omitted for privacy reasons. 
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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 RFP Site Profile as a Basis for Dose Reconstruction 

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

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

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

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

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

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

TBDs. 

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

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

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

and to which objective each particular issue applies. 

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. 
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3.0 ASSESSMENT CRITERIA AND METHODS 

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

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

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

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

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

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

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

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

in SC&A’s site profile review procedure. 

3.1 OBJECTIVES 

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

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

the benefit of the doubt to the claimant. 

3.1.1 Objective 1: Completeness of Data Sources 

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

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

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

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

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

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

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

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

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

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

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

publicly available relating to the RFP 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 which 

occurred at RFP. 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 

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. 
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3.1.3 Objective 3: Adequacy of Data 

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

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

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

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

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

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

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

considered a gap and 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. This 

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

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

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

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

3.1.5 Objective 5: Regulatory Compliance 

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

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

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

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

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

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

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

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

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

accuracy/uncertainty of data. 

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

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

potential internal and external exposure pathways have been considered. 

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

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

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

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

to compensate a claim.

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

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

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

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

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

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

contaminations must be identified. 

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

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

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

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

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

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

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

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

“reasonable estimates” of radiation dose to mean: 

. . . 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 RFP TBDs provide wellorganized and user-friendly information for the dose reconstructor when adequate data were 

available to do that comprehensively.
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ORAUT-TKBS-0011-1, Rev. 00, Technical Basis Document for the Rocky Flats Plant – 

Introduction (Little and Meyer 2004), 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. 

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-0011-2, Rev. 00, Technical Basis Document for the Rocky Flats Plant – Site 

Description (Flack and Meyer 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 RFP TBDs. This document describes SC&A’s review of this section and 

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-0011-3, Rev. 00, Technical Basis Document for the Rocky Flats Plant – 

Occupational Medical Dose (Furman and Lopez 2004), provides an overview of the 

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

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

individual worker dose contributions. The TBD recognizes and draws heavily upon 

assessments at other sites (e.g., the SRP) and the TIB on diagnostic x-rays. SC&A 

reviewed this section for technical adequacy and consistency with other NIOSH TBDs 

and procedures. 

ORAUT-TKBS-0011-4, Rev. 00, Technical Basis Document for the Rocky Flats Plant – 

Occupational Environmental Dose (McDowell-Boyer and Little 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-0011-5, Rev. 00, Technical Basis Document for the Rocky Flats Plant – 

Occupational Internal Dose (Falk et al. 2005), presents background information and guidance to 

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

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

chemical forms of the radionuclides that may have been inhaled by the workers, the 
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recommended assumptions for use in reconstructing internal doses based on lung counts and invitro 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-0011-6, Rev. 00, Technical Basis Document for the Rocky Flats Plant – 

Occupational External Dose (Langsted 2004), 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 applicable 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 and 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 series of individual 

conference calls with NIOSH, ORAU, RFP staff, and SC&A is provided in Attachment 4. 

An extensive comparison was done between the methodologies used in the RFP and other TBDs 

reviewed to date to determine environmental, internal, and external doses. In the case of 

occupational medical exposure, the limited data available warranted additional comparison to the 

SRP and Hanford TBDs. 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 5. 

Prior to the conference call with NIOSH to discuss SC&A questions, SC&A prepared and shared 

with NIOSH a PowerPoint Presentation that provides a summary of the significant findings. The 

presentation, “NIOSH Rocky Flats Plant Site Profile TBDs” is provided in Attachment 6. 

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

findings to finalize the vertical issues2

 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 

project management and if necessary, outside experts, 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 of these have been completed. 

After the Advisory Board and NIOSH have 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 

2

 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 to have a significant impact on dose reconstruction.
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Advisory Board and to NIOSH. We anticipate that, in accord with the procedures followed 

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

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. 
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4.0 SITE PROFILE STRENGTHS 

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

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

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

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

we identified a number of strengths in the RFP site TBDs. These strengths are described in the 

following sections. 

4.1 COMPLETENESS OF DATA 

In general, the site description TBD (ORAUT-TKBS-0011-2) provides an accurate and 

reasonable characterization of the history and operational activities at the RFP for the period of 

production. The major functions of many of the production and support buildings were outlined, 

as well as the major process changes that occurred through the years. There are still some gaps 

related to processing of special materials and recycled uranium. The appendix provides a list of 

documented accidents/incidents in chronological order through November 1976. Detailed 

descriptions were provided for major incidents. This information can assist the dose 

reconstructors in determining whether additional exposure potential exists and whether the total 

dose reconstruction process is claimant favorable. 

In developing the site profile, NIOSH drew upon information from the 415 reports compiled on 

the Site Profile Research Database. These include environmental reports beginning in 1960 that 

present data used for the determination of offsite exposure to the public. Process information 

was drawn from Rocky Flats health studies and technical documents obtained from public 

sources or through the site. NIOSH/ORAU obtained information on the dosimetry program from

site experts knowledgeable in the area. They met with the United Steelworkers on June 23, 

2004, to identify worker concerns and discuss the RFP TBDs. This interaction with workers 

helps provide valuable insight into the site operations and programs. In addition, the issuance or 

planned issuance of several TIBs reflects an ongoing effort by NIOSH to continually improve 

guidance provided to the dose reconstructor. For example, NIOSH is preparing a new ORAUTOTIB-0049 that addresses methodologies for reconstructing dose from Super Type S and other 

insoluble forms of plutonium. This is purported to be an essential new tool to deal with highfired plutonium oxide exposure that often have a much smaller particle size (Falk 2004). 

4.2 ADEQUACY OF DATA 

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

the RFP 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) RFP implemented an environmental monitoring program, including stack monitoring, inperimeter monitoring, offsite monitoring, and groundwater monitoring. 
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(3) Starting in 1964, dosimeters were coupled with security badging, which helped ensure all 

personnel were provided dosimetry. 

(4) NIOSH has provided in sample RFP dosimetry records and an explanation of how to 

interpret these records in both the internal and external TBDs. The inclusion of this type 

of information is commendable and should be continued with future site profiles. 

(5) Figure C-13, presented in Appendix C (Falk 2004), an example of a data report, Health 

Sciences Record, that might be encountered in the claimant’s file. In this example, the 

‘parts per million’ by weight of 241Am was calculated and recorded for the fecal and nose 

smears. 

(6) NIOSH/ORAU have recognized the usefulness of the NDRP study and the job-exposure 

matrix compiled by the University of Colorado. They have recently received data from

the NDRP study and are determining how to incorporate this data into dose 

reconstruction. They are actively pursuing the job-exposure matrix to assist in the 

development of co-worker doses. 

Although RFP has significant quantities of personnel monitoring data, as well as environmental 

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

radiation exposure in the early years of operation. In addition, multiple gaps appear in records of

individuals that were likely monitored. Further records retrieval efforts may be prudent to fill 

some of these gaps. 

4.3 TECHNICAL ACCURACY/CLAIMANT FAVORABILITY 

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

claimant favorability: 

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

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

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

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

for the period of 1951 to 2003 to assist in making the assigned dose claimant favorable. 

The chronological sequences of events at the RFP were outlined in the TBDs, along with 

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

changed over the years. The major areas of external radiation hazards were addressed as 

the functions of the RFP changed over time. 

(2) The dose reconstructor generally uses the solubility that gives the best outcome to the 

claimant. Thus the current information provided in the TBD (Furman and Lopez 2004), 

is no longer valid and will be taken out. 

(3) NIOSH has indicated that all of the interferences, if operative, are claimant favorable, as 

the dose reconstructor uses the recorded lung count data at face value. 
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(4) NIOSH developed a method for handling the variability of recovery and differences in 

the use of the mean value of the MDA or the extreme values. Fifty percent of samples 

have a sample-specific MDA higher than the median MDA for the process. The table on 

the bottom of page 44 gives the median and 5th percentile values of recoveries. The 5th

percentile values of recoveries are an indicator of the variability at the worst-case end of 

the distribution. 

(5) In calculating the total activity in the lung after an incident, NIOSH has developed a 

methodology to account for contributions from both the right and left lung. The DR gets 

the calibration for that period for two detectors, and then multiplies that calibration factor 

by 0.43. The net c/m data for the left chest divided by the adjusted calibration factor 

gives the total americium activity for that count. 

(6) The occupational external dosimetry TBD (Langsted 2004) separated out the different 

types of radiation and their energy ranges to provide more detailed information for dose 

reconstruction (DR). This was not an easy task in view of the many changes in 

operations, radiation fields, dosimetry technologies, and administrative/regulatory 

mandates. A detailed discussion was provided of each type of radiation and its exposure 

potential, dosimetry, dose records, and applicable correction factors for DR. 

(7) ORAUT-OTIB-0027 provides the dose reconstructor with additional information on 

calculating dose from low-energy photons and electrons, and dosimetry uncertainty 

values. 

4.4 CONSISTENCY AMONG SITE PROFILES 

(1) The RFP site profile has introduced a few new concepts not previously used in other site 

profiles. A respirable fraction of 1.0 is applied for environmental dose. This value is 

claimant favorable. 

(2) RFP is the first site profile reviewed to date where the dose reconstructor is directed to 

use both the preceding and following periods to assign dose for gaps in the records. In 

previous site profiles, an unmonitored individual would receive a dose based on the limit 

of detection (LOD) and the exchange frequency. 

(3) The inclusion of a methodology for skin contamination was a positive addition to the site 

profile. This is especially important in facilities that had routine personnel contamination 

incidents. A similar section should be included in other site profiles. 

(4) Although internal uptakes by wounds have been an issue at other sites, the RFP site 

profile is the only site profile reviewed to date that provides direction on assessing 

uptakes via a wound site. Consideration is given to acute and long-term chronic injection 

as a mode of intake. 
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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. 

• 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 (with the one exception being particle size 

where measured values were available for specific operations or incidents).
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5.0 VERTICAL ISSUES 

SC&A has developed a list of key issues regarding the RFP 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: MINIMUM DETECTABLE ACTIVITY 

Suggested use of urine bioassay MDA values for plutonium and americium appear low and are 

likely to yield body burdens/organ doses that are proportionately low and, therefore, claimant 

unfavorable. 

Defining the Issue 

The inhalation of Type S plutonium (and americium) will not only yield very low excretion 

fractions in urine, but because plutonium is a pure alpha emitter, its quantitative detection in 

urine requires complex chemical separation before a sample may either be analyzed by gross 

alpha counting in a gas flow proportional counter or by means of alpha spectrometry that 

employs solid state surface barrier detectors. 

For either method, accurate quantitative measurements require full knowledge of several crucial 

parameters that may include the following: 

(1) The urine volume analyzed relative to a given individual’s 24-hour urine excretion 

volume 

(2) The chemical recovery fraction of the isotope extracted from urine 

(3) Sample absorption, which is defined by the fraction of alphas absorbed (fully or 

partially) that precludes their inclusion in (1) gross alpha counting or (2) the full 

energy peak when alpha spectrometry is employed 

(4) The counting geometry factor and the detector’s counting efficiency (for 2 π

counting geometry, the theoretical upper bound detection efficiency is 50%) 

(5) The statistical uncertainty of converting counts observed in sample to an activity 

value (the magnitude of this uncertainty is defined by the total count time of the 

sample, the value of the blank count rate, and the detector background) 

On page 12 of ORAUT-TKBS-0011-5, the following guidance is provided for dose 

reconstruction: 
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The minimum detectable activity (MDA) for plutonium is presented here for the 

medium conditions. By definition of the medial value, half of the sample-specific 

MDAs are lower than the median value, and half are higher. In most cases the 

dose reconstructor is not likely to have sufficient data to determine the samplespecific MDA, so the median values should be used. [Emphasis added.]

Also included in the guidance are the following median MDA values, as given in 

Table 5.3.1.1.2-1 and reproduced herein: 

Period dpm/24-hr sample

1952 – 1953 0.57 

1954–1962 0.51

1963 0.44

1964 – 1977 0.54 

1978 – 1989 0.24 

1990 – 1992 0.24 

1993 – 0.020 

Note: The unit of the MDA values starting with

1990 is dpm/sample. 

Sample-specific MDA values in the record starting 

1990, if found, should be used instead of the

generic MDA values in this table.

Attachment A of the TBD (pp. 36–37) states that “. . . the general equation for the MDA is 

Equation 6 in the American National Standard, Performance Criteria for Radiobioassay

(HPS 1996),” and presents this equation along with its many terms. Attachment A further states 

that “... Applying this equation to urinalysis methods at Rocky Flats involves determining the 

values of each variable for measurements of the analysis . . . as the methods evolved.” 

Statements of Concern 

SC&A questions the validity of these median MDA values and their use in dose reconstruction 

for the following reasons: 

(1) NIOSH’s ability to quantify critical terms, which define the ANSI equation for 

MDA. In Attachment A of ORAUT-TKBS-0011-5, NIOSH openly states the following: 

. . . the period from 1952 to 1971 for which many of the urinalysis logs 

have been located and analyzed to obtain the information needed to assess 

the MDA . . . urinalysis procedures were primitive and evolving and 

numerous dosimetrically interesting events and intakes were occurring 

at Rocky Flats. [Emphasis added.] 
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NIOSH further implies that these logs/records were far from complete and frequently 

lacked critical information, as given in the following statements: 

... Routine urine samples were typically 24-hr excretions, either one 

continuous 24-hr period (but not taken at the Rocky Flats site) or two 

12-hr periods. Special urine samples could be 24-hr samples, overnight 

samples, or a single voiding. ...The measurement of the sample typically

involved counting the alpha radiation from the processed aliquant of the 

sample and determining the activity of the analyte in the original sample. 

... [for gross alpha counting] Typical counting time was 150 minutes,... 

although from 1952 to 1955 count times of 55, 60, and 75 minutes and in 

1971 count times of 40 and 60 minutes, were also used. A spike sample 

and a reagent blank sample were processed with the workers’ samples, 

sometimes less frequently. The result of the spike sample may or may not

have been used to establish the value of the recovery of the analyte for the 

batch. Similarly, the result of the blank (counts per minute) may or may 

not have been used to establish the value of the blank subtracted from the 

total count rate of the sample. Detector efficiency was stated to be 0.50. 

...No documentation was found concerning the count time used to measure 

the detector background, but the count time is likely to be 150 minutes or 

longer... 

The blank count rate is method specific, and the application of the blank in 

the data analysis was variable between methods and within a method over 

time. The confounder was the intermittent, but persistent, laboratory 

contamination artifacts introduced into blanks and worker samples. 

[Emphasis added.] 

Given the limited data and/or the uncertainties of critical variables that define MDA, it is 

SC&A’s opinion that the use of the median MDA value in instances where bioassay data 

are recorded as zero or background is likely to underestimate actual doses. Included 

among the unsupported assumptions used by NIOSH for deriving median MDA values 

are the following (as given on pg. 46 of the TBD): 

• An assumed sample count time of 150 minutes 

• A theoretical upper-bound detector counting efficiency 

• The assumption that the analyzed sample value represents a full 24-hr urine value 

• The assumption of zero sample self-absorption (i.e., Fa = 1.00) 

• Assumptions regarding the value of So (that incorporate blank sample, detector 

background, etc.) that are not adequately explained/supported 
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(2) The use of median MDA values in dose reconstruction is technically inappropriate 

since urine activities well in excess of median MDA values may not have been 

recorded as explained in the following statements (p. 38 of ORUAT-TKBS-0011-5): 

Tolerance levels were used at Rocky Flats in the 1950s and 1960s as an 

indicator of the maximum permissible amount (activity) of a radionuclide 

excreted per day in a worker’s urine. The technical basis for the values of

tolerance levels used at Rocky Flats has not been identified. The 

significance is that urinalysis results less than 10% of the tolerance level 

were recorded and reported as background (BK on the Urinalysis 

Record Card) or zero, regardless of the underlying sensitivity of the 

method, with some exceptions. [Emphasis added.] 

For plutonium, the required reporting level of $ 0.88 dpm/24-hr is given; and for gross 

alpha, the required reporting level of 8.8 dpm/24-hr is given. This implies that lower 

values were recorded as either zero or background. 

Summary Conclusions 

Based on the information provided in ORAUT-TKBS-0011-5, it is the opinion of SC&A that 

cited median MDA values for Pu, Am, and possibly U are inappropriate for dose reconstruction. 

At a minimum, dose reconstruction should be based on the reporting level of a given analyte. 

A more claimant-favorable approach that is recommended by SC&A is an MDA level that 

incorporates the average of two extreme conditions involving the following four issues: 

(1) A high blank 

(2) Low chemical recovery 

(3) Low sample volume 

(4) Significant sample self-absorption 

As shown on page 47 of ORAUT-TKBS-0011-5, MDA values for two extreme conditions out of 

four yield the following average values: 

Period Average MDA (dpm/24-hr sample)

1952 – 1952 1.55 

1954 – 1962 1.10 

1963 1.30

1964 – 1971 1.61 

5.2 ISSUE 2: OCCUPATIONAL INTERNAL DOSE 

The Occupational Internal Dose TBD, ORAUT-TKBS-0011-5 (Falk 2004), and the Site 

Description TBD, ORAUT-OTIB-0011-2 ((Flack and Meyer 2004), were evaluated to determine 
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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 technical sound assumptions for the calculation of internal doses?

(3) Does the TBD provide guidance for addressing missed internal dose for monitored and 

unmonitored workers?

The Occupational Internal Dose TBD, ORAUT-TKBS-0011-5 (Falk 2004), is presented in four 

general sections─the main body of the report and Attachments A, B and C. The main body of 

the report contains a detailed description of the following. 

• Source term 

• Particle size 

• Solubility 

• In-vitro bioassay methods, minimum detectable activities (MDAs), counting methods, 

reporting levels, and interferences 

• Lung measurements, minimum detectable activities (MDAs), counting methods, 

reporting levels, and interferences 

• Other bioassay data (wound count data, nasal smear and fecal samples) 

• Records and reports 

• Limited information on the assignment of dose to unmonitored workers 

Attachment A describes the minimum detectable activity for urinalysis methods at Rocky Flats. 

Attachment B describes the minimum detectable activity for in-vivo lung counts at Rocky Flats. 

Attachment C provides some examples of records and reports used at Rocky Flats. 

5.2.1 Implications of High-Fired Plutonium in Bioassay Measurements for Acute and 

Chronic Intakes 

The absorption of the inhaled material to the blood depends on its physical and chemical form. 

ICRP recommends that material-specific rates of absorption should be used in the model for 

compounds for which reliable experimental data exist. For other compounds, default values of 

parameters are recommended. At Rocky Flats, high-fired oxides were generated during the two 

big fire accidents, and more than likely as a result of smaller plutonium fires and high 

temperature processes in furnaces, incinerators and production areas. 

Because of these facts, an evaluation of the effect in using the high-fired plutonium (Super S) 

lung retention parameters in the interpretation of bioassay results was carried out using the 

Type S 239Pu lung retention parameters simulating different scenarios (see Attachment 9). The 

four simulated scenarios were: 
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• Chronic inhalation plutonium compounds of different solubility classes with 

urinalysis sampling after 20 years. The total intake was calculated for the 239Pu activity 

of 1 dpm/24-hour in urine samples taken following a chronic inhalation over 20 years of 

plutonium Type S at 1 μm and Type M plutonium compounds, respectively. Equivalent 

doses for the calculated intakes indicates that doses would be higher for the Type M 

compound compared to the insoluble one, except for the large intestine and respiratory 

tract, where the high-fired compounds deliver the highest dose. 

• Chronic inhalation of Type S plutonium compound (15Bq/day) plus acute intake of 

high-fired plutonium oxide with annual urinalysis sampling. A chronic inhalation of 

15 Bq/day of 239Pu Type S, AMAD = 5 μm, through 20 years of exposure with additional 

acute intakes of 10,000 Bq/year, in the beginning of each year, of high-fired 239

Pu 

compound, AMAD = 1μm. Considering this scenario, it is just possible to detect 

measurable 239Pu activity in urine samples just after 10 y of chronic and acute exposure. 

• Chronic inhalation of Type S plutonium compound (100 Bq/day) plus acute intake, 

with quarterly urinalysis sampling. Chronic inhalation of 100Bq/day of 239Pu Type S, 

AMAD = 5 μm, through the 20 years of exposure with additional acute intakes of 

10,000 Bq/year, in the beginning of each year, of Type S 239Pu compound, AMAD = 1μm 

(simulating accidental intake of high-fired Pu). Results indicate that the contribution of 

chronic intake on urinary activity matches and exceeds that from an acute intake of 

10,000 Bq during the 6th year of exposure, and makes it increasingly difficult to detect

acute intakes thereafter unless the measurement system is very sensitive or the intake is 

extremely high. 

• Chronic inhalation of Type S plutonium compound (100 Bq/day) plus acute intake, 

with urinalysis sampling one day after acute intake. Chronic inhalation of 100Bq/day 

of 239Pu Type S, AMAD = 5 μm, through the 20 years of exposure with additional acute 

intakes of 10,000 Bq/year, in the beginning of each year, of Type S 239Pu compound, 

AMAD = 1μm (simulating accidental intake of high-fired Pu). Results show that total 

intake (chronic + acute) is higher than chronic intake alone by only a factor of 1.5, which 

is in the band of normal measurement uncertainty. 

This analysis shows that the incidental acute intake of insoluble plutonium compound, in the first 

20 years, may be difficult to identify because of several factors, including the relatively high 

MDA value (0.01 Bq), the low fraction of activity intake excreted through the urine (10-6 Bq), 

and historic delay in or lack of performing post-incident urinalysis or fecal analysis. It was 

found that the contribution of chronic intake in urinary activity increases over the time of 

exposure, obviating the detection of incidental intakes, unless the activity is extremely high or 

the chronic exposure is very low, or undetectable. 
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5.2.2 Source Term 

The TBD, ORAUT-TKBS-0011-2 (Flack and Meyer 2004), states that neptunium, thorium,

curium and tritium figured in the operations at Rocky Flats. While it is acknowledged in the 

TBD, ORAUT-TKBS-0011-5 (Falk 2004, pg. 7) that site-specific internal dosimetry for 

neptunium, thorium and curium is rare or not available, NIOSH needs to provide some form of 

guidance for approaching dose reconstruction for these radionuclides either in the TBD or in 

supporting OTIBs or workbooks. 

The TBD (Falk 2004) on page 17 states that the workers were exposed to tritium:

Workers were monitored for possible tritium exposures only for special projects or 

situations, starting in 1973. 

The TBD does not give sufficient information on bioassay measurements for tritium as well as 

the form of tritium to which workers were exposed. 

5.2.3 Particle Size 

The TBD (Falk 2004) assumes a 5 μm AMAD as the default particle size for Rocky Flats dose 

reconstructions. Particle sizes and distributions are not available for work areas or incidents, 

except in limited cases. The TBD states on page 9 that in an incident involving a plutonium fire 

in Buildings 776 and 777, Mann and Kirchner (1967) measured a mass median diameter of 

0.3 μm, similar to that reported for an americium fire incident from the same reference. On 

page 9, item 5.2.2.2, instructions on assumed particle sizes for exposure to Am in cases of fire 

are incomplete, in that no value is provided for “x.x.”

Based on direction provided by 42 CFR Part 82, a default 5μm AMAD particle size is only 

applicable in cases where there is no information on particle sizes available. There is some

facility-specific particle size information available. SC&A recommends that the particle sizes 

should be reviewed, and the probability of exposure to particles sizes smaller than or larger than 

5μm should be calculated before adopting the default parameter of 5μm AMAD. 

In cases where the dose is estimated based on air concentration data, the default 5μm AMAD 

particle size is not claimant favorable, as shown in Table 1. This table shows the dose 

coefficients for 1-year equivalent dose, assuming a single intake of 1 Bq of 239Pu by inhalation. 

The comparison between the solubility Type M and Type S shows that independent of the 

particle size, the dose coefficients for Type M are always bigger than for Type S. On the other 

hand, the comparison between the particle sizes AMAD 5 μm and 1 μm shows that independent 

of the solubility, the dose coefficients for AMAD 1 μm are always bigger than for AMAD 

5 μm. 
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Table 1. Dose Coefficients for 1-Year Equivalent Dose Assuming a Single Intake of 1 Bq of 

Plutonium-239 by Inhalation 

AMAD 5 μm AMAD 1 μm Ratio TypeM/TypeS Ratio 1μm /5μm 

Organs Type S Type M Type S Type M AMAD 5 

μm 

AMAD 1 

μm Type S Type M 

Adrenals 1.21E-09 5.82E-08 2.01E-09 7.74E-08 4.81E+01 3.85E+01 1.66E+00 1.33E+00

Bladder Wall 1.21E-09 5.82E-08 2.01E-09 7.74E-08 4.81E+01 3.85E+01 1.66E+00 1.33E+00

Bone Surface 5.17E-07 2.53E-05 8.52E-07 3.35E-05 4.89E+01 3.93E+01 1.65E+00 1.32E+00

Brain 1.21E-09 5.82E-08 2.01E-09 7.74E-08 4.81E+01 3.85E+01 1.66E+00 1.33E+00

Breasts 1.21E-09 5.82E-08 2.01E-09 7.74E-08 4.81E+01 3.85E+01 1.66E+00 1.33E+00

Esophagus 1.21E-09 5.82E-08 2.01E-09 7.74E-08 4.81E+01 3.85E+01 1.66E+00 1.33E+00

St Wall 1.75E-09 5.87E-08 2.36E-09 7.77E-08 3.35E+01 3.29E+01 1.35E+00 1.32E+00

SI Wall 2.56E-09 5.94E-08 2.88E-09 7.82E-08 2.32E+01 2.72E+01 1.13E+00 1.32E+00

ULI Wall 9.41E-09 6.58E-08 7.28E-09 8.21E-08 6.99E+00 1.13E+01 7.74E-01 1.25E+00

LLI Wall 2.52E-08 8.05E-08 1.75E-08 9.11E-08 3.19E+00 5.21E+00 6.94E-01 1.13E+00

Colon 1.62E-08 7.21E-08 1.17E-08 8.60E-08 4.45E+00 7.35E+00 7.22E-01 1.19E+00

Kidneys 1.17E-08 5.30E-07 1.97E-08 7.18E-07 4.53E+01 3.64E+01 1.68E+00 1.35E+00

Liver 8.60E-08 4.24E-06 1.41E-07 5.59E-06 4.93E+01 3.96E+01 1.64E+00 1.32E+00

Muscle 1.21E-09 5.82E-08 2.01E-09 7.74E-08 4.81E+01 3.85E+01 1.66E+00 1.33E+00

Ovaries 5.16E-09 2.54E-07 8.50E-09 3.35E-07 4.92E+01 3.94E+01 1.65E+00 1.32E+00

Pancreas 1.21E-09 5.82E-08 2.01E-09 7.74E-08 4.81E+01 3.85E+01 1.66E+00 1.33E+00

Red Marrow 5.25E-08 2.57E-06 8.65E-08 3.40E-06 4.90E+01 3.93E+01 1.65E+00 1.32E+00

ET Airways 2.61E-05 1.16E-05 1.38E-05 6.21E-06 4.44E-01 4.50E-01 5.29E-01 5.35E-01

Lungs 2.65E-05 1.86E-05 3.96E-05 2.69E-05 7.02E-01 6.79E-01 1.49E+00 1.45E+00

Skin 1.21E-09 5.82E-08 2.01E-09 7.74E-08 4.81E+01 3.85E+01 1.66E+00 1.33E+00

Spleen 1.21E-09 5.82E-08 2.01E-09 7.74E-08 4.81E+01 3.85E+01 1.66E+00 1.33E+00

Testes 5.26E-09 2.58E-07 8.65E-09 3.41E-07 4.90E+01 3.94E+01 1.64E+00 1.32E+00

Thymus 1.21E-09 5.82E-08 2.01E-09 7.74E-08 4.81E+01 3.85E+01 1.66E+00 1.33E+00

Thyroid 1.21E-09 5.82E-08 2.01E-09 7.74E-08 4.81E+01 3.85E+01 1.66E+00 1.33E+00

Uterus 1.21E-09 5.82E-08 2.01E-09 7.74E-08 4.81E+01 3.85E+01 1.66E+00 1.33E+00

 

5.2.4 Solubility 

The TBD recommends that the dose reconstructor should assume that radionuclides are soluble 

for the purpose of dose assessment of the systemic organs and tissues. 

5.2.4.1 Plutonium 

The TBD (Falk 2004) states on page 8 that: 

Plutonium in chemical-processing operations can be either soluble (type M), 

insoluble (type S), or a mixture of solubilities. A claimant favorable approach is 

to assume insoluble plutonium if the qualifying cancer is of the respiratory system 

and to assume soluble plutonium for all other cases. Lung count data in 

conjunction with urine data may help to determine absorption type. 
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The document ORAUT-PROC-0003 (Brackett 2003) states, on page 5, that: 

Because of these differences, “worst-case” assumptions are not necessarily the 

same as those that would result in the largest committed effective dose. They will 

depend on a variety of factors, including radionuclide, organ of interest, and 

latency period of the cancer. For example, while the assumption of absorption 

type S material will generally maximize the lung dose, the assumption of type F 

may result in larger doses to other organs because the material will be removed 

from the lung more rapidly and distributed throughout the body. The process is to 

be claimant-favorable whenever there are unknown values. 

In Table 2, and in Figures 1 and 2, an example of 1-year chronic intake of 239

Pu is provided. The 

committed equivalent doses were based on an activity of 1 Bq in a 24-hour urine sample 

collected in a working day, on the last month of the 1st year of exposure. The committed 

equivalent doses are significantly higher for most of the gastrointestinal organs for the Type S 

compounds compared to those of Type M. It indicates that the claimant-favorable approach will 

depend on the type of cancer, as stated in the document ORAUT-PROC-0003 (Brackett 2003), 

guidance that is not applied in this TBD. 
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Table 2. Comparison of Committed Equivalent Doses for Plutonium-239 Type M and 

Type S Compounds per Bq Plutonium-239 Present in the 24-Hour Working Day Urine 

Sample, Collected at the End of 1-Year Exposure 

Committed equivalent dose 

Organs Inhalation Type M Inhalation Type S 

Sv/Bq excreted Sv/Bq excreted

Ratio 

TypeM/TypeS 

Adrenals 3.59E-03 2.77E-03 1.29E+00

Bladder Wall 3.59E-03 2.77E-03 1.29E+00

Bone Surface 1.48E+00 1.13E+00 1.31E+00

Brain 3.59E-03 2.77E-03 1.29E+00

Breasts 3.59E-03 2.77E-03 1.29E+00

Esophagus 3.59E-03 2.77E-03 1.29E+00

St Wall 3.65E-03 5.72E-03 6.37E-01

SI Wall 3.74E-03 1.01E-02 3.69E-01

ULI Wall 4.55E-03 4.73E-02 9.62E-02

LLI Wall 6.42E-03 1.33E-01 4.83E-02

Colon 5.33E-03 8.43E-02 6.32E-02

Kidneys 3.72E-02 2.95E-02 1.26E+00

Liver 2.44E-01 1.86E-01 1.31E+00

Muscle 3.59E-03 2.77E-03 1.29E+00

Ovaries 1.47E-02 1.12E-02 1.31E+00

Pancreas 3.59E-03 2.77E-03 1.29E+00

Red Marrow 1.50E-01 1.14E-01 1.31E+00

ET Airways 9.96E-01 7.65E+01 1.30E-02

Lungs 2.10E+00 1.21E+02 1.74E-02

Skin 3.59E-03 2.77E-03 1.29E+00

Spleen 3.59E-03 2.77E-03 1.29E+00

Testes 1.49E-02 1.14E-02 1.31E+00

Thymus 3.59E-03 2.77E-03 1.29E+00

Thyroid 3.59E-03 2.77E-03 1.29E+00

Uterus 3.59E-03 2.77E-03 1.29E+00

Years of Exposure to 239Pu: 1 year 

Collection of 24-hr working day urine sample: last month, of the 1st year 

Equivalent Doses calculated for the 1st year after the beginning of work

Excretion of 239

Pu in urine entirely due to inhalation exposure of Type M material (M) – 5 μm AMAD 

Excretion of 239

Pu in urine entirely due to inhalation exposure of Type S material (S) – 5 μm AMAD 
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One Year Committed Equivalent Dose for Pu-239 (Sv per Bq

excreted in the 24hr working day urine samples after 1 year of

work)

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 Marro

w

Skin

SpleenTestes

Thymus

Thyroid

Uterus

Organs

C

o

m

mitted Equi

v

alent

D

o

s

e (

S

v/B

q in urin

e)

Inhalation Type M

Inhalation Type S

Figure 1. 1-Year Committed Equivalent Dose for All Organs due to 1-Year Chronic

Inhalation of Plutonium-239 Type M and Type S Compounds 

One Year Committed Equivalent Dose for Pu-239 

(Sv per Bq excreted in the weekday urine sample after 1 year

dr

u

Br

St

SI

ULI W

LLI Wa

Colo

idney

Live

Muscl

Ovarie

ancreas

ro

Ski

SpleenTeste

Thymu

Thyro

Uter

C

o

m

mitted Equi

v

ale

ne)

of work)

1.00E+01

nt

1.00E-03

1.00E-02

1.00E-01

1.00E+00

enals

 Wall

rface

Brain

easts

ophagus

 Wall

Wall

all

ll

n

s

r

e

s

w

n

s

s

id

us

D

ose (

Sv/

B

q i

n uri

Inhalation Type M

Inhalation Type

A

Bladder

Bone S

Es

K

P

Red Mar

Systemic organs

S

Figure 2. 1-Year Committed Equivalent Dose for Systemic Organs due to 1-Year Chronic

Inhalation of Plutonium-239 Type M and Type S Compounds 
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There is an uncertainty in the definition of the uranium compound. The TBD (Falk 2004) states 

that: 

y cases, the compo anium involved in an intake is not identified. 

no Rocky Flats fic data for e ranium. In g

chemical proce s occurred as type M or as mixtures of type M 

ype S. Reconstructio use the most claimant favorable mixture. For 

that occur in meta areas, the c vorable assu

f the qualifying ca s in the respiratory system. For other cancer 

the more claimant fa ssumption is type M. Reconstructions should 

default value of 5

The values presented in Table 3 i hat the comm ivalent doses icantly 

higher for most of the gastrointestinal organs for the ura pounds Type S compared to 

the Type M. It means that the cla orable appro epend on the ancer as

stated on the document ORAUT-PROC-0003 (Brackett 2003), but not applied in this TBD. 

Figu ally repre able 3 comm valent dose.

5.2.4.2 Uranium 

In man und of ur

There are site-speci nriched u eneral,

intakes in ssing area

and t ns should

intakes

S i

l-working laimant fa mption is

type ncer site i

sites, vorable a

use the μm.

llustrate t itted equ are signif

nium com

imant-fav ach will d type of c

res 3 and 4 graphic sent the T itted equi
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Table 3. 1-Year Uranium-234 Committed Equivalent Doses per Bq Uranium-234 Present 

in the 24-Hour Working Day Urine Sample, Collected at the End of 1-Year Exposure 

Type M Type S Ratio 

Sv/Bq (excreted) Sv/Bq (excreted) TypeM/TypeS

Adrenals 1.79E-05 1.61E-05 1.11E+00

Bladder Wall 1.91E-05 1.71E-05 1.12E+00

Bone Surface 1.99E-03 1.79E-03 1.11E+00

Brain 1.79E-05 1.61E-05 1.11E+00

Breasts 1.79E-05 1.61E-05 1.11E+00

Esophagus 1.79E-05 1.61E-05 1.11E+00

St Wall 2.03E-05 8.49E-05 2.39E-01 

SI Wall 2.36E-05 1.88E-04 1.25E-01 

ULI Wall 5.25E-05 1.06E-03 4.95E-02 

LLI Wall 1.19E-04 3.06E-03 3.88E-02 

Colon 8.09E-05 1.92E-03 4.21E-02 

Kidneys 1.89E-03 1.71E-03 1.10E+00

Liver 7.98E-05 7.14E-05 1.12E+00

Pancrea

Red Marrow 2.09E-04 1.87E-04 1.11E+00

ET Airways 3.69E-02 1.83E+00 2.01E-02 

1.61E-05 1.11E+00

Spleen 1.79E-05 1.61E-05 1.11E+00

Te

Th

Th 0

Ut 00

Yea

Collection of 24-hr working day urine sample: last month, of the 1st year 

u

Muscle 1.79E-05 1.62E-05 1.11E+00

Ovaries 1.79E-05 1.62E-05 1.11E+00

s 1.79E-05 1.61E-05 1.11E+00

Lungs 7.00E-02 2.59E+00 2.70E-02

Skin 1.79E-05

stes 1.79E-05 1.61E-05 1.11E+00

ymus 1.79E-05 1.61E-05 1.11E+00

yroid 1.79E-05 1.61E-05 1.11E+0

erus 1.79E-05 1.61E-05 1.11E+

rs of Exposure to 234U: 1 year

Eq ivalent Doses calculated for the 1st year after the beginning of work
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One Year Committed Equivalent Dose for U-234 (Sv per Bq

excreted in 24 hr working day urine samples after 1 year of

work)

1.00E+00

1.00E+01

rine)

1.00E-05

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

C

D

1.00E-04

1.00E-03

1.00E-02

m

mitted E

q

u

se (

Sv/

B

q i

n

1.00E-01

o iv

ale

nt

o u

Inhalation Type M

Inhalation Type S

Figure 3. 1-Year Committed Equivalent Dose for All Organs due to 1-Year Chronic 

Inhalation of Uranium-234 Type S, Type M and Type S Compounds 

One Year Committed Equivalent Dose for U-234 (Sv per Bq

excreted in 24 hr working day urine samples after 1 year of

work) 

1.00E

1.00E

Adrenals

Bladder Wall

Bone Surface

Brain

Breasts

Esophagus

St Wall

SI Wall

Kidneys

Liver

Muscle

Ovaries

Pancreas

Red Marro

w

Skin

SpleenTestes

ic organs

C

o

m

mitte

d E

q

uivalen

D

o

s

e (

S

v/B

q in urin

t

1.00E-02

e)

-05

-04

1.00E-03

ULI

LLI

Co

 Wa

 Wal

lo

l

n

l

l

Thy

Thy

Ut

mus

roid

erus

System

Inhalation Type M

Inhalation Type S

Figure 4. 1-Year Committed Equivalent Dose for Systemic Orga 1-Year Chronic

Inhalation of Uranium-234 Type S, Type M and Type S Compounds 

ns due to
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SC&A finds that solubility should be assigned based on the type of cancer, as recommended in 

the 42 CFR Part 82, which 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 

5.2.5 Bioassay Data – Uncert

 the earlier tim

u an u m es higher than the more recent values, as 

hown i b .1 - e 2 s .

e sam i h n e t o h

ults o n T e c e s A

may result in a wide variation. The TBD 0 t A e es

n p p h o i

c o s

v l r p p s o

p a l i

h i o

page k t a

Not until 19

d 2p k

t t c o i

c b a , v b e

n s i r o h

rec s e o l re t s h s

iges r t h e i e s s e

prep p d e t fi e a h

imp s h a 2 h m e

b v x f s m o

i 2 A m e r y p

eek. T e o st e e p e o

1⁄2 m D y B t

f a o s e a a

sam c in d e lt 8 T o

just h n w i li o h

existing knowledge of the material and process. 

ainties and Minimum Detectable Activities

The TBD (Falk 2004) shows that in es, the m

20 tim

edian MDA values for determination 

of P

s

d Am in rine sa ples are about 30 and 

n the Ta les 5.3.1 .2-1 and 5.3.1.2.2 1 on pag s 12 and 14 (Falk 004), re pectively

Th

s

ple-spec fic MDA may be igher tha the gen ric ones abulated r described with t e

re f the uri alyses. his is du to the fa t that th recovery strongly influence the MD

and (Falk 20 4), in At achment on pag 39, stat

that: 

Dependi g on the rocess, s iked samples, samples to w ich a kn wn activ ty of

the analyte was added, were generally processed with ea h batch f sample . The

recovery alues ca culated f om the s iked sam les were the ratio of the c unt

rate of spiked sam le to the verage count rate of four to six samp es depos ted on

the planc et or plate with m nimal pr cessing. 

On 52 (Fal 2004), i states th t:

73 were some plutonium samples spiked with an internal tracer (first 

236

Pu an , later, 24 Pu). All lutonium samples were spi ed with an internal 

racer af er 1978. Experien e has sh wn that a significant variab lity of recovery

an exist within a atch of s mples. Therefore the reco ery of a atch spik does

ot nece sarily ind cate the ecovery f each sample in t e batch.

The overy is trongly d pendent n severa factors lated to he analy is of eac sample, uch

as d tion of o ganic ma erial of t e sampl , compos tion of th sample , reagent , and car in

the

than the tab

aration o

ulated v

f the sam

alues. 

le. Base on thes factors, he speci c-sampl MDA m y be hig er

The ortant is ue that s ould be ddressed in this TBD (Falk 004) is t e assess ent of th

missed dose. Ta le 4 pro ides an e ample o the asse sment of issed d se considering a

20-year chronic ntake of 39

Pu, Type S, AM D = 5 μ , during ight hou s per da , 5 days er

w he miss d doses f r the fir 20 cons cutive y ars of ex osure was estimat d, based n

the edian M A value of urinal sis, MDA = 0.57 dpm (0.01 q). The daily intake and to al

intake or the ye rs of exp sure are hown on Table 4. The valu s were c lculated ssuming that

the ple was ollected the last ay of th year and the resu was 0.2 5 dpm. able 5 sh ws 

that intakes igher tha 104 Bq ould be detected n the ear er times f Rocky Flats. T e
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NIOSH develop guidance for use b ssess the missed dose, especially 

for the insoluble compounds, since a small amount in urine may represent a significant intake 

ose. The same applies for pecially for 

an ). 

ble a o r ic ti u 2 e

5 s e ia s d

y a ( t

estimated missed equivalent doses are shown on Table 5. SC&A strongly recommends that 

y the dose reconstructor to a

and d uranium; the MDA was considerably high, es

uranium compounds Type S (UO2 d U3O8

Ta 4. Estimated Int ke due t 20-Yea s Chron Inhala on of Pl tonium- 39 (Typ S,

AMAD μm) Ba ed on th 1⁄2 Med n MDA Value of Urinaly is (0.285 pm)

Time ( ears) D ily intake Bq) Total In ake (Bq)

1 9.72E+01 2.43E+04

2 4.85E+01 2.43E

E

E

E

E

E

E

E

0 E

11 1.02E+01 2.81E+04

14 8.46E+00 2.96E+04

15 8.04E+00 3.01E+04

+04

3 3.24E+01 2.43 +04

4 2.45E+01 2.45 +04

5 1.99E+01 2.49 +04

6 1.69E+01 2.53 +04

7 1.46E+01 2.56 +04

8 1.31E+01 2.62 +04

9 1.19E+01 2.68 +04

1 1.09E+01 2.72 +04

12 9.45E+00 2.84E+04

13 8.91E+00 2.90E+04

16 7.68E+00 3.07E+04

17 7.38E+00 3.14E+04

18 7.08E+00 3.18E+04

19 6.77E+00 3.21E+04

20 6.58E+00 3.29E+04

Table 5. M he 1⁄2 Median

Missed equivalent dose (Sv) x Time of exposure (years) 

issed Doses for the Organs due to Inhalation of Plutonium-239, Based on t

MDA Value of Urinalysis 

Organs 1 2 3 4 5 6 7 8 9 10 

Adrenals 1.30E-05 3.46E-05 6.19E-05 9.31E-05 1.28E-04 1.66E-04 2.04E-04 2.48E-04 2.93E-04 3.38E-04

Blad

Wall 3.38E-04

Bone

Surf -02 4.41E-02 6.25E-02 8.30E-02 1.04E-01 1.29E-01 1.55E-01 1.82E-01 

Brain 1.30E-05 3.45E-05 6.19E-05 9.31E-05 1.28E-04 1.66E-04 2.04E-04 2.48E-04 2.93E-04 3.38E-04 

Brea 05 9.31E-05 1.28E-04 1.66E-04 2.04E-04 2.48E-04 2.93E-04 3.38E-04 

Esophagus 1.30E-05 6.19E-05 9.31E-05 1.28E-04 1.66E-04 2.04E-04 2.48E-04 2.93E-04 3.38E-04 

St W 2.6 E-04 1.42E-04 1.81E-04 2.19E-04 2.63E-04 3.08E-04 3.53E-04 

SI Wall 4.68E-05 2.03E-04 2.41E-04 2.86E-04 3.31E-04 3.77E-04 

ULI 5.77E-04

LLI 1.04E-03

der

1.30E-05 3.46E-05 6.19E-05 9.31E-05 1.28E-04 1.66E-04 2.04E-04 2.48E-04 2.93E-04

ace 5.29E-03 1.51E-02 2.83E

sts 1.30E-05 3.46E-05 6.19E3.45E-05 

all 5E-05 4.82E-05 7.55E-05 1.07

6.89E-05 9.62E-05 1.28E-04 1.64E-04 

Wall 2.17E-04 2.41E-04 2.71E-04 3.04E-04 3.44E-04 3.87E-04 4.28E-04 4.77E-04 5.27E-04

Wall 6.10E-04 6.40E-04 6.74E-04 7.13E-04 7.62E-04 8.13E-04 8.58E-04 9.21E-04 9.82E-04
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Table 5. Missed Doses for the Organs due to Inhalation of Plutonium-239, Based on the 1⁄2 Median 

MDA Value of Urinalysis 

Missed equivalent dose (Sv) x Time of exposure (years) Organs 10 

Colon E-04 7.76E-04

1 2 3 4 5 6 7 8 9 

3.87E-04 4.12E-04 4.44E-04 4.80E-04 5.23E-04 5.71E-04 6.13E-04 6.68E-04 7.22

Kidneys 1.37E-04 3.09E-04 4.96E-04 6.93E-04 8.96E-04 1.10E-03 1.30E-03 1.51E-03 1.73E-03 1.93E-03

Liver 3.52E-02

Muscle 04 3.38E-04

Ovaries -03 2.09E-03

Pancreas E-04 3.38E-04

Red 

Marrow 1.53E-03 2.83E-03 4.34E-03 6.05E-03 7.91E-03 9.77E-03 1.19E-02 1.40E-02 1.62E-02 

ET .70E+00

Lun 9.92E-01

Skin -04

Sple 04

Test 03

Thy 3.38E-04

Thy 3.38E-04

Uter E-04

Time of exposure (years) 

8.71E-04 2.56E-03 4.89E-03 7.79E-03 1.12E-02 1.52E-02 1.94E-02 2.44E-02 2.96E-02

1.30E-05 3.46E-05 6.19E-05 9.33E-05 1.28E-04 1.66E-04 2.04E-04 2.48E-04 2.93E5.27E-05 1.53E-04 2.91E-04 4.63E-04 6.67E-04 9.02E-04 1.15E-03 1.44E-03 1.76E

1.30E-05 3.45E-05 6.19E-05 9.31E-05 1.28E-04 1.66E-04 2.04E-04 2.48E-04 2.93

5.37E-04

Airways 3.56E-01 6.26E-01 8.42E-01 1.02E+00 1.17E+00 1.30E+00 1.40E+00 1.52E+00 1.62E+00 1

gs 5.56E-01 6.40E-01 6.97E-01 7.45E-01 7.92E-01 8.38E-01 8.70E-01 9.16E-01 9.57E-01

1.30E-05 3.45E-05 6.19E-05 9.31E-05 1.28E-04 1.66E-04 2.04E-04 2.48E-04 2.93E-04 3.38E

en 1.30E-05 3.46E-05 6.19E-05 9.31E-05 1.28E-04 1.66E-04 2.04E-04 2.48E-04 2.93E-04 3.38Ees 5.36E-05 1.56E-04 2.96E-04 4.70E-04 6.79E-04 9.19E-04 1.17E-03 1.47E-03 1.80E-03 2.13Emus 1.30E-05 3.45E-05 6.19E-05 9.31E-05 1.28E-04 1.66E-04 2.04E-04 2.48E-04 2.93E-04

roid 1.30E-05 3.45E-05 6.19E-05 9.31E-05 1.28E-04 1.66E-04 2.04E-04 2.48E-04 2.93E-04

us 1.30E-05 3.45E-05 6.19E-05 9.31E-05 1.28E-04 1.66E-04 2.04E-04 2.48E-04 2.93E-04 3.38

O

13 14 15 16 17 18 19 20 

Adrenals 3.89E-04 4.35E-04 4.86E-04 5.41E-04 5.96E-04 6.52E-04 7.12E-04 7.72E-04 8.26E-04 8.95E-04 

rgans 11 12

Bladder Wall 3.89E-04 4.35E-04 4.86E-04 5.41E-04 5.96E-04 6.52E-04 7.12E-04 7.72E-04 8.26E-04 8.

 Surfac

95E-04

Bone e 2.13E-01 2.41E-01 2.74E-01 3.07E-01 3.42E-01 3.78E-01 4.15E-01 4.53E-01 4.89E-01 5.33E-01

Brain 3.89E-04 4.35E-04 4.86E-04 5.41E-04 5.96E-04 6.52E-04 7.12E-04 7.72E-04 8.26E-04 8.95E-04 

Breasts .26E-04 8.95E-04

Esophagus -04 8.95E-04

St Wall -04 9.15E-04

SI Wall -04 9.41E-04

ULI Wall -03 1.18E-03

LLI Wall 60E-03 1.67E-03 1.75E-03

Colon 

Kidn E-03

Live -01

Mus 8.95E-04

Ova 5E-03

Panc 04

Red E-02

ET A E+00

Lun 3E+00

Skin 95E-04

Sple -04

Test -03

Thy -04

Thy 9E-04 4.35E-04 4.86E-04 5.41E-04 5.96E-04 6.52E-04 7.12E-04 7.72E-04 8.26E-04 8.95E-04 

Uter s 3.89E-04 4.35E-04 4.86E-04 5.41E-04 5.96E-04 6.52E-04 7.12E-04 7.72E-04 8.26E-04 8.95E-04 

3.89E-04 4.35E-04 4.86E-04 5.41E-04 5.96E-04 6.52E-04 7.12E-04 7.72E-04 8

3.89E-04 4.35E-04 4.86E-04 5.41E-04 5.96E-04 6.52E-04 7.12E-04 7.72E-04 8.26E

4.05E-04 4.51E-04 5.03E-04 5.58E-04 6.13E-04 6.70E-04 7.31E-04 7.93E-04 8.46E

4.29E-04 4.75E-04 5.28E-04 5.85E-04 6.39E-04 6.97E-04 7.60E-04 8.14E-04 8.73E

6.35E-04 6.84E-04 7.42E-04 8.03E-04 8.60E-04 9.22E-04 9.89E-04 1.05E-03 1.11E

1.11E-03 1.17E-03 1.24E-03 1.31E-03 1.37E-03 1.45E-03 1.53E-03 1.

8.40E-04 8.92E-04 9.53E-04 1.02E-03 1.09E-03 1.15E-03 1.22E-03 1.29E-03 1.35E-03 1.43E-03 

eys 2.15E-03 2.34E-03 2.57E-03 2.78E-03 3.00E-03 3.22E-03 3.45E-03 3.65E-03 3.85E-03 4.09

r 4.16E-02 4.76E-02 5.44E-02 6.18E-02 6.91E-02 7.68E-02 8.49E-02 9.35E-02 1.01E-01 1.11E

cle 3.89E-04 4.35E-04 4.86E-04 5.41E-04 5.96E-04 6.52E-04 7.12E-04 7.72E-04 8.26E-04

ries 2.47E-03 2.84E-03 3.24E-03 3.68E-03 4.12E-03 4.58E-03 5.08E-03 5.58E-03 6.05E-03 6.6

reas 3.89E-04 4.35E-04 4.86E-04 5.41E-04 5.96E-04 6.52E-04 7.12E-04 7.72E-04 8.26E-04 8.95EMarrow 1.86E-02 2.08E-02 2.32E-02 2.56E-02 2.81E-02 3.06E-02 3.32E-02 3.57E-02 3.80E-02 4.09

irways 1.80E+00 1.86E+00 1.93E+00 2.01E+00 2.07E+00 2.14E+00 2.21E+00 2.27E+00 2.31E+00 2.38

gs 1.04E+00 1.07E+00 1.10E+00 1.14E+00 1.17E+00 1.21E+00 1.24E+00 1.27E+00 1.29E+00 1.3

3.89E-04 4.35E-04 4.86E-04 5.41E-04 5.96E-04 6.52E-04 7.12E-04 7.72E-04 8.26E-04 8.

en 3.89E-04 4.35E-04 4.86E-04 5.41E-04 5.96E-04 6.52E-04 7.12E-04 7.72E-04 8.26E-04 8.95E

es 2.52E-03 2.88E-03 3.30E-03 3.75E-03 4.20E-03 4.67E-03 5.18E-03 5.69E-03 6.16E-03 6.78E

mus 3.89E-04 4.35E-04 4.86E-04 5.41E-04 5.96E-04 6.52E-04 7.12E-04 7.72E-04 8.26E-04 8.95E

roid 3.8

u
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The TBD (Falk 2004, pg. 16) states that gross alpha urinalysis was used for workers who were 

potenti ing

period:

e

for

 or plutonium (after 1963). Interferences are likely, because the 

ethods were non-specific. Isotopes are all alpha-emitting isotopes of the 

d to other alpha emitters, the TBD should provide some guidance 

evaluate the internal dose. There is no bioassay data or air concentration data to allow the 

ose reconstruction for neptunium, curium and thorium. NIOSH should provide a claimantfavorable guidance document. 

5.2.6 Lung Count Data 

5.2.6.1 Americium and Plutonium 

The TBD (Falk 2004, pg. 18) states that there are two sources of interferences to consider: 

The first is the 63-keV gamma doublet of 234

Th from depleted uranium operations 

being mistaken for 241

Am for lung counts with the NaI or phoswich detector 

systems. This interference was most troublesome for workers with residual lung 

depositions of plutonium and americium who subsequently worked in depleted 

uranium operations. The second interference is the contribution of count from 

241Am not in the lungs, for example, contributions from contamination on the skin, 

from material being cleared from the upper respiratory system, or from ingested 

material. A positive detection of 241

Am did not necessarily indicate an intake of 

the plutonium/americium mixture, especially for a lung count in response to an 

incident. 

For the second interference, the skin contamination contribution may be true for accidents or 

incidents, but not for the material being cleared from the upper respiratory system or from

ingested material, since the transit time in the upper compartments of the human alimentary tract 

is fast. The reference values, for an adult male, of transit time for the luminal contents in the oral 

cavity is 15 seconds, esophagus is 8 seconds and stomach is 1.5 hr. The contribution of the 

upper region of the gastrointestinal tract in the lung count should not be considered. If this 

assumption was true, it was indicative that an inhalation intake occurred indeed. This statement 

is not correct. 

ally exposed to uranium, plutonium, and other alpha emitters in the same monitor

Gross alpha measurement is a non-specific analysis used for workers who wer

potentially exposed to both uranium and plutonium in the same monitoring 

period. Workers who were potentially exposed to other alpha-emitting 

radionuclides, such as neptunium and curium, may also have been monitored

gross alpha. Urinalysis methods are discussed in Attachment 5A. The gross 

alpha method was discontinued in the early 1970s, likely 1973. The results are 

reported as dpm/24-hr sample of either enriched uranium (the default analyte 

through 1963)

m

analyte. 

Since the workers were expose

to

d
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he TBD shows some inconsistencies regarding the assumption of the date of intake to assess 

the dose from lung count data. In the TBD (Falk 2004, pg. 20), it is stated that: 

The assu ption of the intake dat raightforward and sho ce

maximizing the plutonium lung deposition (intake date is close to of the

lung count) and maximizing the accrued lung dose (intake date is far from the 

date of the lung count). In addit hoice of intake date for the lung count 

data sho d be coordinated with he associated urine dat

According to th BD, the in-vivo lung nitoring was performed for workers who were 

exposed, or had the potential to be exposed to airborne plutonium. It is not clear if it was applied 

as part of the routine monitoring progra as special monitoring applied in incidental 

intakes. A better guidance document should be established, especially to ow to define

the date of inta n order to have a clai rable approach. The lu data associated

with urine data is not the best way to evaluate the occurrence of an acute intake. In general, the 

compounds involved in the incidental intakes were high-fired plutonium m, which are 

both very insoluble compounds. Unless the intake was very high, it would be very difficult to 

detect incidental intakes using urine data, since the activity expected to be found in urine 

excretion is very low, in the order of 10 intake), while the activity expected to be found 

in feces excretion is in the order of 10-2 (Bq/Bq intake), in the first days after intake, as shown on 

Table 6. 

he TBD (Falk 2004, pg. 18) states the following:

Reporting levels are not easily defined, because quantification was preceded by 

d workers

were classified as positive unknowns or some variation. Starting in 1974, the 

There i ctors should choose the appropriate value 

f ppm Am, and also there is no clear explanation of the basis to choose 1,000 ppm as the 

5.2.6.2 Uncertainties 

T

m e is not st uld balan

the date

ion, the c

ul that for t a.

e T count mo

m or just

clarify h

ke i mant-favo ng count

or uraniu

-6 (Bq/Bq

T

verification counts and professional judgments. Before 1974, the practice was 

not to quantify a positive detection of 241Am unless the deposition could be 

associated with a known incident with a known ppm 241Am. Affecte

practice was changed to quantify the plutonium depositions for positive 

unknowns by assuming a default value of 1,000 ppm 241Am on the date of the

most likely intake or on the date of the first positive lung count. The ppm 241Am 

was then calculated for the date of the lung count to account for the ingrowth of

241Am from the nuclear transformation of 241Pu and the radioactive decay of the 

initial 241Am. 

s no clear instruction on how the dose reconstru

241 o

default value for ppm 241Am.
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m-239

/Bq intake) for 1 μm AMAD

Table 6. Intake Fraction in Urine and Feces after the Inhalation of 1 Bq of Plutoniu

Type S Compound (Bq

Inhalation 239Pu – Type S -1 μm Days after intake 24-hr Urine (Bq/Bq intake) 24-hr Feces (Bq/Bq intake)

1 2.03E-06 6.10E-02

2 1.23E-06 8.85E-02

3 8.04E-07 4.65E-02

4 6.09E-07 2.01E-02

5 5.02E-07 8.45E-03

6 4.35E-07 3.83E-03

7 3.90E-07 2.05E-03

8 3.59E-07 1.37E-03

9 3.38E-07 1.11E-03 

10 3.23E-07 1.00E-03 

12 3.05E-07 9.14E-04 

20 2.87E-07 7.61E-04 

30 2.84 14E-04 

4.09E-04 

37E-04 

2.83E-07 2.81E-04 

2.84E-07 2.36E-04 

90 2.84E-07 2.00E-04 

2.85E-07 1.71E-04 

14 2.95E-07 8.68E-04

16 2.91E-07 8.30E-04

18 2.88E-07 7.95E-04

E-07 6.

40 2.84E-07 4.99E-04

50 2.83E-07

60 2.83E-07 3.

70 

80 

100 

Thorium/Depleted Uranium: 

The TBD states that one of the methodologies used to assess the internal exposure to depleted 

uranium was to estim 238 234 ate the U activity based on the Th measurement in lung, assuming 

equilibrium. In the TBD (Falk 2004, pg. 21), the following is mentioned: 

s

t

has not been assessed.

ll

The major uncertainty is the assumption of equilibrium of the 234Th with the 238U

before 1990, when depleted uranium was still being processed. Part of the 

process was to remove decay chain radionuclides, especially thorium, by heating

the uranium ingot to drive the smaller atoms of thorium to the surface or top of 

the ingot, which was then cut off. The result was depleted uranium metal with a 

deficiency of 234Th for several weeks plus scrap depleted uranium with an exces

of 234Th (supe-requilibrium). Super-equilibrium is claimant favorable. The effec

of a deficiency of 234

Th

The assumption of full equilibrium is not necessarily claimant favorable. If the worker is 

exposed to depleted uranium metal with a deficiency of 234Th, the assumption of equilibrium wi

underestimate the 238U activity in the lung of the workers exposed to depleted uranium metal 

with a deficiency of 234Th. 
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tamination, 

specially after the advent of the wound counter in 1957. The TBD (Falk 2004) states on page 

22 that: 

Wound count infor ction. The

relevant items are the urin f the mode and date of

intake, and whether there was residual plutonium at the wound site. If there was 

residual plutonium at the wound site, the dose reconstructor should consider an 

acute injection into the blood stream plus a possible long-term chronic injection. 

The profile of the urine data following the date of the wound provides guidance on 

the proportion of the acute and chronic components. If there was no detected 

residual plutonium at the wound site, there would have been an acute injection 

into the blood stream. 

Since no appropriate wound model is available currently, the approach is claimant favorable for 

most types of cancer. This approach is not appropriate for some special cases like lymph node 

and skin cancers. For lymph nodes, a special approach should be proposed. It is known that in 

cases of wound contamination with 241

Am, the largest fraction of the Am activity measured in 

the body is associated with tissues near the puncture wound. Popplewell and Ham (1989) have 

The TBD (Falk 2004) states, on page 21, that the MDA for 238U has not been determined 

rigorously, and it is based on the 241

Am data: 

The MDA for 238U has not been determined rigorously. However, the 238U 

worker-specific MDA can reasonably be expected to be a multiple of the 241Am 

worker-specific MDA because the detected photons (63 keV and 59.5 keV) are 

very close in energy. As described in Section 5.3.2.2.1 for using the calibration 

factor for 241

Am to determine the 238U activity, the 238U worker-specific MDA can 

be obtained by multiplying the 241

Am worker-specific MDA by 9.4. That result is 

divided by 0.89 to obtain the worker-specific MDA for depleted uranium. (As 

noted in Section 5.3.2.1.2 for americium and plutonium, MDA values are reported 

on report forms 1995 and later, but are not worker-specific. The dose 

reconstructor should disregard these MDA values.)

The determination of the MDA is very important since it may be high; the low measurable 

activity value may represent a significant dose. NIOSH should provide MDA values for 238U 

and 235U. 

The TBD (Falk 2004) does not consider the 235U measurements in lungs to evaluate the enriched

uranium exposure for the period before 1995, as stated on page 21. 

5.2.7 Other Bioassay Data 

5.2.7.1 Wound Count Data 

According to the TBD, wound contamination was very likely to happen. Any wound that 

occurred in a work area involving plutonium was monitored for plutonium con

e

mation is largely irrelevant to dose reconstru

alysis data, the identification o
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escribed a case of wound contamination 18 years after a puncture wound to the left hand, which 

Am activity (a product of Pu decay) was associated with the left arm axillary lymph nodes 

and left hand, 12% was measured easured in the liver. Guilmette 

nd Durbin (2003) have proposed a classification of wound retention for radionuclides. The 

retention patterns, which are related to the chemical properties of the radionuclides, were 

grouped empirically into four categories, as shown in Figure 5. The criterion for grouping the 

radionuclides was based on the fraction of injected activity remaining at the wound site 1 d after 

deposition, and the rate(s) at which the retained fraction subsequently cleared from the site. The 

authors have classified Am3+ and Pu4+ (at doses < 0.2μg) as strongly retained elements. The 

characteristic of this group is such that these two radioisotopes are slowly cleared from the 

wound site. Plutonium is classified as an avidly retained element. Besides the chemical 

properties of the radionuclides, an important variable affecting wound-site retention is the 

deposited mass. Guilmette and Durbin (2003) have reported an example, at about 1 week after 

injection, 34% of 239Pu with an initial mass of 11 ng was retained. This can be compared to 

51%–66% retention for masses from 0.10 to 1.0 μg, and 96% for a 16 μg mass. At >25 μg, 90% 

retention was observed at the end of 1 month after injection, and 72% at 2 months. Thus, the 

pattern of increasing retention with increased injected Pu mass is consistent with the attainment 

of higher mass concentrations within the wound site. The wound model for strongly and avidly 

retained elements, like Am and Pu, is not available at the moment, but a draft of the model for 

wound-site retention of soluble radionuclides has been proposed by Guilmette and Durbin 

(2003). The wound biokinetic model consists of five compartments comprising the wound site, 

and blood and lymph node compartments, which receive radionuclides that are being cleared 

from the wound site. The compartmental design is based on the physical and chemical properties 

of the deposited radioactive material, which can be soluble, a mixture of soluble and colloidal 

material, particulate or fragment, as shown in the Figure 6. All the reported facts show that a 

claimant-favorable approach needs to be developed for lymph node cancers. 

d

had resulted in the deposition of a splinter of plutonium metal. Eighty percent of the measured 

241 241

in the skeleton, and 1% was m

a
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Figure 5. Default Retention G

Depo

roups (weak, moderate, strong and avid) for Radionuclides 

sited in Intramuscular Wound Sites

(Guilmette and Durbin 2003) 

Figure 6. Schematic Compartmental Representation of the Draft NCRP Wound Biokinetic 

Model 

(Guilmette and Durbin 2003) 
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5.2

Fecal samples were taken when incidents occurred. The fecal data are very important to evaluate 

any incidental intak ble co As s able 6 er to d

incidental intakes using fecal data rather than urine data. The activ d to

ur tion is d

to in feces is on t of 10 ) t days ke.

SC&A suggests that val an f feca ments ata are in

the individual claima onal fi FP bio ords) b ized f

reconstruction from xposu

5. -worker D gnme

ORAUT-TKBS-0011-5 (Falk 2004) does not provid e for ting th r

un red worker D is i e in th

5.3 SSUE 3: PO AL FO STIO SURE AY

Th ion pathw ternal has n consid he TBD. It is well 

kn ough n is, in the ma ay of i estion ur.

SC ends tion o , am and ur mpoun

in art of al dos I trac

5. utonium Example 

Th t of pluto enters through ingestion is more important in terms of bioassay 

results and doses per Bq excreted than through inhal r exam worke re

ha 1 year, the ommitted equivalent doses to the diffe ns, per

excreted in a 24-hr working day urine sam

Ta

A t be

ig all others, the doses due to 

in d 8 graphically represent the

in

.7.2 Fecal Samples 

e of insolu mpounds. hown in T , it is easi

ity expecte

etect

be found in

ine excre very low; on the order of 10-6 (Bq/Bq intake). -2

In contrast, the activity expecte

be found excretion he order (Bq/Bq intake in the firs after inta

the retrie d review o l measure (if that d available

nt’s pers le or in R assay rec e emphas or dose

episodic e res.

2.8 Co ose Assi nt

e guidanc reconstruc e dose fo

monito s. The TB ncomplet is regard. 

I TENTI R INGE N EXPO PATHW

e ingest ay for in exposure ot been ered in t

own that alth inhalatio general, in pathw ntake, ing may occ

&A recomm

as a p

that inges f plutonium ericium, anium co ds be

cluded the intern e for the G t.

3.1 Pl

e amoun nium that

ation. Fo ple, if a r’s exposu

d lasted 1-year c rent orga Bq 239Pu

ple at the end of the working year, is illustrated in 

ble 7.

s can be seen from Table 7, doses coming from the 239

Pu ingestion pathway should no

nored, except for the organs related to the respiratory tract. For

gestion are higher than the ones due to inhalation. Figures 7 an

formation provided in Table 7. 
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Table 7. 1-Year Plutonium-239 Committed Equivalent Doses per Bq Plutonium-239 

Present in 24-Hour Working Day Urine Sample, Collected at the End of 1-Year Exposure 

Inhalation 

Type M 

Inhalation 

Type S 

Ingestion

f1=0.0005 

Ingestion

f1=0.0001 

Ingestion

f1=0.00001

Organs Sv/Bq excreted Sv/Bq excreted Sv/Bq excreted Sv/Bq excreted Sv/Bq excreted

Adrenals 3.59E-03 2.77E-03 4.17E-03 4.17E-03 4.17E-03 

Bladder Wall 3.59E-03 2.77E-03 4.17E-03 4.17E-03 4.19E-03 

Bone Surface 1.48E+00 1.13E+00 1.73E+00 1.73E+00 1.73E+00

Brain 3.59E-03 2.77E-03 4.17E-03 4.17E-03 4.17E-03 

Breasts 3.59E-03 2.77E-03 4.17E-03 4.17E-03 4.17E-03 

Esophagus 3.59E-03 2.77E-03 4.17E-03 4.17E-03 4.17E-03 

St Wall 3.65E-03 5.72E-03 2.03E-02 8.49E-02 8.12E-01 

SI Wall 3.74E-03 1.01E-02 4.45E-02 2.06E-01 2.02E+00

ULI Wall 4.55E-03 4.73E-02 2.48E-01 1.23E+00 1.22E+01

LLI Wall 6.42E-03 1.33E-01 7.18E-01 3.56E+00 3.56E+01

Colon 5.33E-03 8.43E-02 4.50E-01 2.23E+00 2.23E+01

Kidneys 3.72E-02 2.95E-02 4.24E-02 4.24E-02 4.24E-02 

Liver 2.44E-01 1.86E-01 2.86E-01 2.86E-01 2.86E-01 

Muscle 3.59E-03 2.77E-03 4.17E-03 4.17E-03 4.19E-03 

R

ET Airways 9.96E-01 7.65 4.17E-03 4.17E-03 

Lungs 2.10E+00 1.21E+02 4.17E-03 4.17E-03 4.17E-03 

Skin 3.59E-03 2.77E-03 4.17E-03 4.17E-03 4.17E-03 

Spleen 3.59E-03 2.77E-03 4.17E-03 4.17E-03 4.17E-03 

Testes 1.49E-02 1.14E-02 1.76E-02 1.76E-02 1.76E-02 

Thymus 3.59E-03 2.77E-03 4.17E-03 4.17E-03 4.17E-03 

Thyroid 3.59E-03 2.77E-03 4.17E-03 4.17E-03 4.17E-03 

Uterus 3.59E-03 2.77E-03 4.17E-03 4.17E-03 4.22E-03 

Ovaries 1.47E-02 1.12E-02 1.72E-02 1.73E-02 1.79E-02

Pancreas 3.59E-03 2.77E-03 4.17E-03 4.17E-03 4.17E-03

ed Marrow 1.50E-01 1.14E-01 1.76E-01 1.76E-01 1.76E-01

E+01 4.17E-03 

Years of Exposure to 239Pu: 1 year 

Collection of 24-hr working day urine sample: last month, of the 1st year Equivalent Doses calculated for the 

1st

 year after the beginning of work 

Excretion of 239

Pu in urine entirely due to inhalation exposure of Type M material (M) – 5μm AMAD 

Excretion of 239

Pu in urine entirely due to inhalation exposure of Type S material (S) – 5μm AMAD 

Excretion of 239

Pu in urine entirely due to ingestion exposure (f1 = 5 x10-4) 

Excretion of 239

Pu in urine entirely due to ingestion exposure (f1 = 1x10-4) 

Excretion of 239

Pu in urine entirely due to ingestion exposure (f1 = 1x10-5) 
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One Year Committed Equivalent Dose for Pu-239 (Sv per Bq excreted after 

1.00E-03

1.00E-02

1.00E-01

1.00E+01

Adr

adder

e Su

n

Breasts

Esophagus

St Wall

SI Wall

ULI Wall

LLI Wall

Colon

Pancreas

Red Marrow

ET Airways

Lungs

Skin

Spleen

Com

mitted Equival v

weekend urine samples after 1 year of work)

1.00E+02

1.00E+03

/Bq in Urine

Inhalation Type M

1.00E+00

ent Dose S

Inhalation Type S

Ingestion f1=0.0005

Ingestion f1=0.0001

Ingestion f1=0.00001

enals

 Wa

rfac

Brai

Bl

Bon

ll

e

Kid

M

Ov

ney

Live

uscl

arie

s

r

e

s

Te

Th

Th

Ut

ste

ymu

yroi

erus

s

s

d

Organs

One Year Committed Equivalen Pu-239 (Sv per d after

weekend urine s ter 1 year of wor

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

1.00E+02

BlR

er

Com

mitted Equivalent Dose Sv/Bq in 

Urine

t Dose for Bq excrete

amples af k)

Adr

adder

one Surf

Br

Breas

Esopha

St

SI

ULI

LLI

Co

Kidney

Li

Mu

Ova

Pancr

ed Mar

SplTe

Thy

Thy

U

enals

 W

t

all

ace

ain

ts

gus

 Wall

 Wall

 Wall

 Wall

lon

s

ver

scle

ries

eas

row

Skin

een

stes

mus

roid

B

Systemic Organs

us

Inhalation Type M

Inhalation Type S

Ingestion f1=0.0005

Ingestion f1=0.0001

Ingestion f1=0.00001

Figure 7. 1-Y mitted Equivalent r Organs due to 1-Year Chronic Intake of 

Plutonium- per Bq Excreted our Working ne Samples after 

1 Year of Work 

ear Com Dose fo

239, in Sv in 24-H Day Uri

Figure 8. 1-Year Committed Equivalent Dose for Systemic Organs due to 1-Year Chronic 

Intake of Plutonium-239, in Sv per Bq Excreted in 24-Hour Working Day Urine Samples

after 1 Year of Work 
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5.3.2 Americium Example 

For americium, the activity that enters the body via ingestion is more important than the activity 

that enters the body via inhalation 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 241

Am excreted in the 24-hr working day urine sample at the end of the 

working year, is illustrated in Table 8. 

Table 8. 1-Year Americium-241 Committed Equivalent Doses per Bq Americium-241 

Present in 24-Hour Working Day Urine, Collected at the End of 1-Year Exposure 

Organs Inhalation Type M Ingestion (f1=0.0005) 

Adrenals 5.18E-04 5.82E-04 

Bladder Wall 5.18E-04 6.33E-04 

Bone Surface 3.92E-01 4.50E-01 

Brain 5.16E-04 5.75E-04 

Breasts 5.16E-04 5.75E-04 

Esophagus 5.18E-04 5.75E-04 

St Wall 5.43E-04 6.86E-03 

LLI Wall 2.77E-01 

Colon 1.22E-03 1.73E-01 

Kidneys 1.17E-02 1.31E-02 

Liver 1.48E-01 1.69E-01 

Muscle 5.16E-04 5.95E-04 

Ovaries 6.50E-03 7.74E-03 

Pancreas 5.18E-04 5.93E-04 

Red Marrow 3.28E-02 3.75E-02 

ET Airways 3.72E-01 5.75E-04 

Lungs 8.72E-01 5.77E-04 

Skin 5.16E-04 5.79E-04 

Spleen 5.16E-04 5.86E-04 

Testes 5.18E-04 5.82E-04 

Thymus 5.18E-04 6.33E-04 

Thyroid 3.92E-01 4.50E-01 

Uterus 5.16E-04 5.75E-04 

SI Wall 5.80E-04 1.63E-02 

ULI Wall 9.02E-04 9.50E-02 

1.64E-03

Years of Exposure to 241Am: 1 year

Collection of the 24-hr working day urine sample: last month, of the 1st year 

Equivalent Doses calculated for the 1st year after the beginning of work

Excretion of 241

Am in urine entirely due to inhalation exposure of Type M 

material (M) - 5μm AMAD 

Excretion of 241

Am in urine entirely due to ingestion exposure (f1 = 5 x10-4) 

As can be seen from Table 8, doses due to the 241Am ingestion pathway should not be ignored. 

Figures 9 and 10 graphically represent the information provided in Table 8. 
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One Year Committed Equivalent for Am-241

1.00E-04

1.00E-01

Adrenals

Bladder Wall

B ace

 W

 Wall

LLI Wall

lon

e

Re row

ET Airways

Lungs

ro

erus

C

o

m

mit q se

(Sv per Bq excreted in 24hr working day urine samples after 1 year of work)

1.00E+00

 S

v/Bq in

1.00E-02

uivalent D

o

Urine

Inhalation Type M

1.00E-03

ted E

one Surf

Br

s

St

SI

UL

Br

I

ain

easts

ophagus

Wal

E

l

all

Co

Kidn

Mus

Ov

Pancr

d Mar

ys

r

e

s

eas

Live

cl

arie

Skin

leen

stes

ymus

id

SpTe

Th

Thy

Ut

Organs

Ingestion f1=0.0005

One Year Commi lent fo

(Sv per Bq n 24hr work ne sample ar of work

00E-04

1.00E-03

1.00E-01

00E+00

Adrenals

adder Wall

e Surface

SI Wall

ULI Wall

LLI Wall

Ovaries

Pancreas

Red Marrow

Skin

Uterus

tted Equiva r Am-241

 excreted i ing day uri s after 1 ye )

Co

1.

a

1.00E-02 le

rin

1.

Bq

Bl

Bon

Bra

eas

hagu

St W

in

ts

s

all

Br

Esop

Col

idney

Liv

Muscl

on

s

er

e

K

SpleTest

ym

hyr

e

o

n

es

us

id

Th

T

Systemic organs

m

mitted Equiv

in U

nt

e)

 Dose (

Sv/

Inhalation Type M

Ingestion f1=0.0005

Figur omm ivalent Organ -Year C ntake of

Americium-241, in Sv per Bq Excreted our Wo y Urine s after

1 Year of Work 

 

Figure onic

Intak ples

e 9. 1-Year C itted Equ Dose for s due to 1 hronic I

in 24-H rking Da Sample

10. 1-Year Committed Equivalent Dose for Systemic Organs due to 1-Year Chr

e of Americium-241, in Sv per Bq Excreted in 24-Hour Working Day Urine Sam

after 1 Year of Work 
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5.3.3 Uranium Example 

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 in 24-hr working day 

urine sample, is illustrated in Table 9. 

As can be seen from Table 9, doses coming from the 234

U ingestion pathway should not be 

ignored, especially in the case of cancer to the gastrointestinal (GI) tract. Figures 11 and 12 

graphically represent the information provided in Table 9. 

Table 9. 1-Year Uranium-234 Committed Equivalent Doses per Bq Uranium-234 Present 

in 24-Hour Working Day Urine, Collected at the End of 1-Year Exposure 

Inhalation 

Type M 

Inhalation 

Type S 

Ingestion

f1=0.02 

Ingestion

f1=0.002 

Sv/Bq excreted Sv/Bq excreted Sv/Bq excreted Sv/Bq excreted

Adrenals 1.79E-05 1.61E-05 1.92E-05 1.92E-05 

Bladder Wall 1.91E-05 1.71E-05 2.05E-05 2.05E-05 

Esophagus 1.79E-05 1.92E-05 1.92E-05 

St Wall 2.03E-05 8.49E-05 3.59E-05 1.86E-04 

SI Wall 2.36E-05 1.88E-04 5.98E-05 4.35E-04 

ULI Wall 5.25E-05 1.06E-03 2.67E-04 2.54E-03 

LLI Wall 1.19E-04 3.06E-03 7.44E-04 7.38E-03 

Colon 8.09E-05 1.92E-03 4.71E-04 4.62E-03 

Kidneys 1.89E-03 1.71E-03 1.99E-03 1.99E-03 

Liver 7.98E-05 7.14E-05 8.72E-05 8.72E-05 

Muscle 1.79E-05 1.62E-05 1.92E-05 1.92E-05 

Ovaries 1.79E-05 1.62E-05 1.92E-05 1.95E-05 

Pancreas 1.79E-05 1.61E-05 1.92E-05 1.92E-05 

Red Marrow 2.09E-04 1.87E-04 2.25E-04 2.25E-04 

ET Airways 3.69E-02 1.83E+00 1.92E-05 1.92E-05 

Lungs 7.00E-02 2.59E+00 1.92E-05 1.92E-05 

Skin 1.79E-05 1.61E-05 1.92E-05 1.92E-05 

Spleen 1.79E-05 1.61E-05 1.92E-05 1.92E-05 

Testes 1.79E-05 1.61E-05 1.92E-05 1.92E-05 

Thymus 1.79E-05 1.61E-05 1.92E-05 1.92E-05 

Thyroid 1.79E-05 1.61E-05 1.92E-05 1.92E-05 

Uterus 1.79E-05 1.61E-05 1.92E-05 1.92E-05 

Bone Surface 1.99E-03 1.79E-03 2.14E-03 2.14E-03 

Brain 1.79E-05 1.61E-05 1.92E-05 1.92E-05 

Breasts 1.79E-05 1.61E-05 1.92E-05 1.92E-05 

1.61E-05

Years of Exposure to 234U: 1 year

Collection of the 24-hr working day 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 M material (M)-5μm AMAD 

Excretion of 234

 U in urine entirely due to inhalation exposure of Type S material (S) - 5μm AMAD 

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) 
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One Year Committed Equivalent Dose for U-234 

(Sv per Bq excreted in 24 hr working day urine samples after 1 year of work)

1.00E-05

Co

1.00E-04

s

all

e

in

ts

gus

all

all

all

all

on

s

er

e

es

eas

R ow

ET s

in

es

us

id

us

Organs

m

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

mitted Equivalent Dose (Sv/Bq in

urine)

Inhalation Type M

Inhalation Type S

Ingestion (f1=0.02)

Ingestion (f1=0.002)

Adrenal

Bladder W

Bone Surfac

Bra

Breas

Esopha

St W

SI W

ULI W

LLI W

Col

Kidney

Liv

Muscl

Ovari

Pancr

ed Marr

 Airway

Lungs

Sk

SpleenTest

Thym

Thyro

Uter

One Year Committed Equivalent Dose for U-234

(Sv per Bq excreted in 24 hr working day urine samples after 1 year of work)

1.00E-01

1.00E-05

1.00E-04

Adrenals

Bladder Wall

one Surface

Brain

Breasts

sophagus

St WallSI Wall

ULI Wall

LLI Wall

Colon

Kidneys

Liver

Muscle

Ovaries

Pancreas

ed Marrow

Skin

Spleen

Testes

Thymus

Thyroid

Uterus

C

ommitted E

1.00E-03

1.00E-02

quivalent D

ose (

S

v

urine)

B

E

R

Systemic organs

/B

q in

Inhalation Type M

Inhalation Type S

Ingestion (f1=0.02)

Ingestion (f1=0.002)

Figure 11. 1-Year Committed Equivalent Dose for Organs due to 1-Year Chronic Intake of

34, in Sv per Bq Excreted in 24-Hour Working Day Urine Samples after 1 Year 

ic

Uranium-2

of Work 

Figure 12. 1-Year Committed Equivalent Dose for Systemic Organs due to 1-Year Chron

Intake of Uranium-234, in Sv per Bq Excreted in 24-Hour Working Day Urine Samples 

after 1 Year of Work 
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enarios, the amount of 

lutonium, americium, and uranium that enters through ingestion is very important in terms of 

inte e

use of bioassay results to back-calc roduce higher doses for certain 

rgans, if dose contributions from the ingestion pathway of exposure are considered, instead of 

just the inhalation pathway. SC&A finds that it is not claimant favorable to ignore the ingestion 

athway for internal contamination. The ingestion of insoluble compounds should be included 

as a part of the internal dose, especially in cases of cancer in organs of the gastrointestinal tract. 

The TB her than

uranium

5.4.1 

The TB adequate assessment of potential doses associated with recycled 

Flack and Meyer

at contains

dioisotopes not found in virgin uranium, i.e., only 238U, 235

U, and 234U. Recycled uranium

n y

236U, and 

traces of certain fission products and transuranic radionuclides. 

at

aterial. For instance, according to a DOE 2000 report, DOE’s Idaho National 

ngineering Laboratory (INEEL) shipped 219 kilograms of 236

U recovered from previously 

irradiat n be

signific hed.”

(DOE 2

Further

5,403,0 re of

deplete ounts 

of recy veral

contam

It is apparent that the ingestion pathway of exposure was not given consideration in the 

derivation of co-worker data. As demonstrated by the preceding intake sc

p

rpretation of bioassay results and conversion to organ equivalent doses per Bq excreted. Th

ulate intake and doses will p

o

p

5.4 ISSUE 4: OTHER RADIONUCLIDES 

D is incomplete in its review of the historic dose contribution of radioisotopes ot

, plutonium, and americium. 

Inadequate Information Regarding Recycled Uranium 

D does not provide an

uranium, which is reflected in the following comment: 

While TBD reviewer comments indicate that Paducah was processing U 

beginning in 1953, available Rocky Flats records do not indicate whether fission 

product or TRU contaminated U was processed at Rocky Flats. (

2004) 

Recycled uranium is so called because it is recovered from reprocessing plants after it has 

already been through a reactor one or more times. This creates uranium th

ra

contai s all three of these radioisotopes, as well as other isotopes of uranium, notabl

The TBD authors (and SC&A) were unable to find records at Rocky Flats pertaining to recycled

uranium. Such information appears to exist based on a review of data provided by the sites th

generated this m

E

ed reactor fuel in 1955 to the RFP (DOE 2000). The radiological dangers of 236

U ca

ant as it “can result in significant gamma fields when secular equilibrium is approac

000, pg. iv).

more, according to the 2000 report done by INEEL, Rocky Flats shipped

00 pounds of recycled uranium in the form of billets to INEEL for the manufactu

d uranium armament (DOE 2000). In 1999, 60 samples were taken from equal am

cled uranium sent to INEEL from Rocky Flats and Fernald, which found se

inants, including transuranics and fission products (see Table 10). 
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Tab om

Rocky Flats and Fernald 

le 10. Radionuclide Contaminants in Depleted Uranium Shipped to INEEL fr

In light of the following conclusion reached in a 1985 DOE review, the values in the above tab ma

not be representative of the contaminants in recycled uranium sent to Rocky Flats over the 30-yea

period in which this material was processed:

Informal specifications in the form of “gentlemen’s agreements’ did evolve and 

h

le y

r

ave been used over time... A formal, technically sound, understood and accepted 

d either within or between

Docum en’s

agreem or

stanc taff-level agreements between sites and within the RFP 

ere si

The ab e

measur y for

worker

5.4.2 

In 1996 of

metals OE

assessm

ity

ampered emergency egress conditions, fire protection 

specification for maximum transuranic and fission product contaminants in 

uranium recycle material has probably never existe

sites.... (DOE 1985) 

 

entation or the lack thereof regarding the nature of these informal staff-level “gentlem

ents,” and their changes over more than 30 years should be addressed by NIOSH. F

in e it is quite possible that early s

w gnificantly less stringent than later “gentlemen’s agreements.” 

sence of data regarding the levels of contaminants in recycled uranium, workplac

ements for contaminants in recycled uranium, and the lack of individual dosimetr

s processing recycling uranium needs to be addressed by NIOSH. 

Inadequate Information Regarding Highly-Enriched Uranium Storage 

Vulnerabilities 

, the RFP held 7.7 metric tons of Highly-Enriched Uranium (HEU) consisting mostly

in the form of pits, parts, samples, and scrap (DOE 1996). According to a 1996 D

ent: 

In addition to the existing facility condition vulnerabilities previously identified in 

the Plutonium Vulnerability Assessment... the most significant Highly Enriched

Uranium Environment, Safety and Health Vulnerabilities involved critical

safety weaknesses, h

program weaknesses, authorization basis deficiencies, and institutional 

weaknesses... Criticality safety vulnerabilities were identified in Buildings 707, 
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he TBD needs to more completely address this aspect of the operational history of Rocky Flats. 

5.4.3 

The TB early

1980s, Federal nuclear complex

as made by the Atomic Energy Commission and its successor agencies to produce 233U and to 

records indicate that the RFP routinely handled kilogram quantities of 23 U. 

t U 

ere made into uranyl nitrate solutions at Hanford at a concentration of approximately 

Processing operations involving 233U were carried out at RF starting in 1965 and 

rious uranium

compounds, conversion to metal, casting, metal fabrication and waste and residue 

itrate solution was

transferred to receiving tanks. Fluoride precipitation was then used to remove 

the

uranium was converted to peroxide. The peroxide was shipped to Building 881 

ing operations took

place in Building 881, while other fabrication steps were handled in Building 883. 

ere collected, treated, packaged, and shipped to various locations 

ff-site. Uranium oxides and green salt residues were converted to uranyl nitrate 

e

and machining chips were burned to oxide in Building 881 and 

776/777, 881 and 886.. Fissile material and fissile material storage areas were 

not adequately labeled or lacked any labeling of limits or mass loadings. 

..Additionally of the 126 criticality safety limit infractions identified since 1993, 

only 40 have been closed out... Poor housekeeping, inappropriate drum storage, 

or improperly roped-off radiation boundaries presented life safety issues.... (DOE 

1996, pg. ES-1) 

T

Inadequate Information Regarding the Processing of Uranium-233 

D makes only a passing reference to the processing of 233U. From 1945 to the

a considerable amount of effort involving several sites in the

w

develop military and civilian applications for this fissile material. Between 1965 and the mid 

1980s DOE 3

From he mid-1960s to the early 1970s during peak production, hundreds of kilograms of 233

w

300 grams of 233U per liter. Uranium-233 uranyl nitrate solutions were then sent to the Oak 

Ridge site, at a rate of 20 to 40 kilograms per shipment. Oak Ridge subsequently shipped 233U 

uranyl nitrate to the RFP in Colorado, and possibly other facilities, for processing (Hanford 

1968). 

According to a history of 233U processing at Rocky Flats: 

ending in 1982. Activities included chemical processing of va

disposal... (RFP 1999, pg. 7) 

233U processing began in Building 771 where the uranyl n

‘hot’ (highly “radioactive) daughter products (primarily Th-228) and the 

where it was calcined to an oxide, which in turn was converted to uranium 

tetrafluoride (UF4), and reduced to 23% metal using a thermal reaction. The 

metal was then cast into feed ingots, which were in turn recast into pieces from 

which the final parts were fabricated. Casting and machin

Final component assembly and inspection occurred in Building 777. All wastes

and residues w

o

solution in Building 771 and returned to OR in the original shipping casks. Som

casting skulls
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bsequently converted to a nitrate solution along with the other oxides. 

Of particular concern is exposure to U contaminants. Uranium-232 is co-produced with 233U 

onstitutes a potentially significant external hazard for U handling and processing. For 

i , t duction of 233U at Hanford, the RFP 

rocessed U in uranyl nitrate and then fabricated U metal components in kilogram 

The first processing operations at the Rocky Flats Plant involving uranium-233 

The TB

operati

5.5 

ome areas of potential exposures re not sufficiently covered, in 

the P

refe ed

in the d

e addr nal dosimetry TBD and new OTIBs to be issued in the near 

utu .

com le

5.5.1 

was o t there were a number of areas of concerns when evaluating the information 

me of 

nd However, this information is not necessarily consistent

su

(RFP 1999, pg. 5)

232

by irradiation of thorium and is considerably more radioactive than 238

U. This is due to highenergy gamma radiation emitted in the decay scheme of 232U daughter products (228thorium, 244radium, and 228thalium). Typically, 233U currently stored at DOE sites contains 5 to 50 parts

per million of 232U (DOE 1996). Although 232U concentrations are small, its gamma radiation 

233 c

nstance he year following the initiation of large-scale pro

233 233 p

quantities (DOE 2000a). According to a DOE contractor assessment done in 2000: 

(U-233) occurred in 1965. ...The material also contained approximately 50 parts 

per million (ppm) contaminant ...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]. (DOE 2000a) 

D needs to address the potential exposure history and dose estimation for 233U in RFP

ons.

ISSUE 5: EXTERNAL DOSE CALCULATION AND METHODOLOGIES 

S that were not addressed, or we

RF TBD, ORAU-TKBS-0011-6 (Langsted 2004) dated January 10, 2004, Rev. 00 (hereafter

rr to as the TBD) could lead to underestimated, or missed, external doses to RFP workers 

ose reconstruction (DR) process. According to ORAUT/NIOSH some of these areas will 

b essed further in a revised exter

f re However, these issues and areas of concern are included in this review for the purpose of 

p teness.

Areas of Concern in Records and Dose Reconstruction 

It bserved tha

for DR provided in the TBD and OTIB. There appeared to be inconsistencies or gaps in so

the information. Some of these areas of concern are as follows: 

5.5.1.1 Dose Reconstruction from Badge Readings versus Historical Documents

The TBD, page 11, and ORAU-OTIB-0027 (Smith 2005), pages 5–7, provide information on 

skin a penetrating radiation doses.

with the information provided in some of the historical records at the RFP (Owen 1964 and 

Putzier 1982). In 1964, Owens defines the RFP doses as follows: 
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Penetrating dose = hard gamma dose + soft gamma dose + 1/3(x-ray dose) + 

RAU-OTIB-0027 (Smith 2005), pages 5–7, uses the open window (OW), cadmium filter (CD), 

ithout relating it to the procedures in 

lace at the time the doses were recorded. In addition, page 10 of the TBD states that further 

e.

deep

gamma dose and the neutron dose were recorded as the total penetrating dose, and not 

ses in the worker’s file. Ruttenber (2003) provides some more

details on the subject, stating that only the total penetrating dose (gamma + neutron) was 

ge

ted

when attempting to separate gamma and 

neutron doses and performing DR. 

e

to

e. It is unclear what this really means and how it will 

affect DR. 

here was a variety of dosimetry methods, dose notations, and record systems at the RFP 

in

Skin dose = hard gamma dose + soft gamma dose + x-ray dose + beta dose + 

neutron dose 

neutron dose 

Hand dose = (hard gamma dose + soft gamma dose + x-ray dose) x 2.5 or 5.0 

depending on area and chemical form. (Owen 1964) 

O

and brass filter (BR) to define skin and penetrating doses, w

p

research is necessary to determine which dosimetry method was used to record a worker’s dos

5.5.1.2 Incomplete or Inconsistent Dose Information 

Some areas in the TBD where information, which would be useful in DR, seems to be lacking or 

appears inconsistent are as follows. 

• Page 11, Section 6.3.1, states that during the early period (around 1952–1975), the

recorded as separate do

recorded for the time periods of 1952–1958 and 1964–1975, and that for 1976, a lar

number of neutron and gamma doses were erroneously recorded, apparently due to 

mistakes made in updating the computer data system. Additionally, Ruttenber (2003,

pg. 25) states that there were many problems with the 1970 neutron dosimetry and rela

records. These items could lead to problems

• Page 14 of the TBD, Section 6.3.2.2.1, states that from July 1984 to October 1984, som

neutron dose was recorded, but the gamma and total dose were zero. This could lead

DR problems if not resolved. 

• Page 14 of the TBD, Section 6.3.2.2.2, describes a possible manual correction needed for

1984–1986, where the recorded deep dose was much greater than the neutron plus 

gamma dose, but ends with the statement that dose components were not provided in the

letter and, therefore, were not mad

T

throughout its years of operations. In view of this, it is of concern that the information used

dose reconstruction correctly reflects the dose received by the worker. It is not apparent from

reading the TBD that the changes that took place during the years in dose 

determination/recording are adequately tracked and accounted for. 
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ess.

and

 was reported as zero dose at the RFP 

tarting in 1993. The DR should take this into account when assigning dose and base it on the 

er the

rk area

d.

her

should be done concerning this issue. 

d

ure. Calculations

how that the worker’s dosimeter worn on the chest would have received approximately 50% of 

ge. This would

ad to a two-fold underestimate of dose. 

5.5.2 Job Exposure Matrix Study 

In Section 6.3.4.3, page 18 of the TBD, it states that DOE funded the Job Exposure Matrix study 

that was conducted by the University of Colorado-Health Services Center. Based on discussions 

with NIOSH and ORAU staff, NIOSH has attempted to gain access to this study without succ

NIOSH agrees that this document is important to characterizing worker exposures at RFP

will be especially critical to external dose reconstruction, where co-worker assignments will 

benefit from job-specific exposure data. 

5.5.3 Below 10 mrem Dose Reported as Zero 

Page 20, Section 6.4.4, states that any dose below 10 mrem

s

appropriate LOD at the time. 

5.5.4 Badges Only Calibrated for One Work Location 

Individuals sometimes worked on temporary or overtime assignments in other facilities (or at 

other jobs) in addition to the assignment for which their badge was calibrated (as stated on page

19, Section 6.4.3 of the TBD). The dosimetry department could not detect this change in 

radiation environment. This may have occurred often enough to create a significant 

underestimate of dose to workers, because of the change in the dosimetry response. Howev

TBD (Langsted 2004, Section 6.6.3.4, pg. 27) assumes that each worker stayed in the wo

for which the badge was calibrated, and therefore no corrections for photon energy were neede

This could lead to the underestimate of the worker’s dose if this was a common practice. 

5.5.5 Exposure Geometry 

In the dose reconstruction process, the assignment of isotropic (ISO) and rotational (ROT) 

instead of AP geometry may not reflect the true radiation dose to some workers; such as 100%

rotational for site support personnel or 50% for support personnel as given in the TBD (Langsted 

2004, Table 6-5, pg. 23). This may not always be claimant favorable in some cases. Furt

investigation

In addition, security guards could have received greater exposures than were recorded, because 

they were responsible for transporting material throughout the site using trucks with cabs an

trailers. The radioactive material on the trailers generated a great deal of external radiation 

exposure at times. Eventually, they had to increase the spacing between the cab and the trailer 

by 6 feet to reduce exposure. The security guards were receiving posterior-anterior exposures

but wore their dosimeters on their chest, which were calibrated for A-P expos

s

the dose incident on the worker’s back for photons in the 60 keV to 1 MeV ran

le
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creased. Page 40, Section 6.8.4.1, addresses beta angular dependence 

nd states that the assembled badge displays severe angular dependence to beta exposure and that 

5.5.6 Angular Dependence for Beta/Gamma, NTA Film, and TLDs 

In this TBD, and other TBDs as well, the issue of angular dependence for different types of 

radiation and different dosimetry systems used through the years is not sufficiently addressed. 

For example, the TBD (Langsted 2004, Section 6.6.4.1, pg. 27) mentions photon angular 

dependence, but seems somewhat vague about the issue. On page 27 it states that the dosimetry

response increases with angle, but on page 28 it states that the Panasonic system response 

decreased as the angle in

a

for TLD badges, it might record only 36% to 59% of the true beta dose at +30 degrees (this could

sult in underestimating the true beta dose, especially to the extremities where the worker’s

r

to

e

st

some buildings and jobs, workers wore their dosimeters outside their protective lead aprons 

ldings/jobs workers wore

eir dosimeters under their protective lead aprons where the dosimeters would not register the 

n the

e 60 keV photons would be transmitted; 

7% of 100 keV photons would be transmitted; 70% of 200 keV photons would be transmitted, 

those working with materials that 

ontained radioisotopes that emitted photons with energies under 100-200 keV and wore their 

n would have little affect on the penetrating dose. 

re

normal movements would not tend to average it out). The Y-12 ORAUT-TKBS-0014-6 

(Kerr 2003, pg. 18) states that the recorded dose of record is likely too low at non A-P angles fo

beta/photon doses. For more uniformity in DR and consistency between sites, NIOSH needs

arrive at a policy to address angular dependence of the different dosimetry systems used at th

AEC facilities/labs including (1) NTA film, (2) beta/photon film, (3) TLND neutron, and 

(4) TLD beta/gamma badges; with modifications for specific locations and time periods. At lea

some general guidelines for consistency would be helpful. 

5.5.7 Use of Dosimeters under Lead Aprons 

In

where the dosimeter would register the exposure dose, but in some bui

th

total dose to the head, neck, and to the lower parts of the body (see Attachment 2). This could 

create a situation where large exposures could accumulate over time without showing up o

dosimetry records. This is especially true for those workers who worked with relatively lowenergy photon emitters (such as 241

Am) around gloveboxes or assembly lines for extended 

periods. Using a lead thickness of 0.5mm (the standard lead-apron thickness available), 

calculations show that only approximately 10% of th

2

and 95% of 500 keV photons would be transmitted. Therefore,

c

dosimeters under their lead aprons could be at risk of having under-recorded doses to the head,

neck, and lower body. A series of studies at the Pantex Plant by Passmore (1995) confirms this 

in that he showed that single-sided 0.5 mm-lead aprons decreased the photon dose in areas of 

stored weapon by 57%, but did not significantly impact the neutron dose readings. Therefore, he

recommended that the dosimeter be worn outside the lead apron for correct photon dose 

monitoring. 

 

During dose reconstruction, these individuals could be identified by their job functions and the 

fact that their dose records should show relatively low skin/shallow dose (because the lowenergy photons were absorbed by the lead apron) compared to a relatively larger penetrating

photon and neutron dose, because the lead apro
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This ar ed

dose, b

5.5.8 

Some tasks required multiple ports of the gloveboxes to be opened (unshielded) during a shift. 

y that

est. For example, if the worker opened the 

lower gloveport while sitting down, and then stood up and started working in an upper gloveport, 

lean over a

ovebox and receive doses to the head and face that would not necessarily be registered by a 

ould be

vestigated because of the possibility of unmonitored doses to gloveport/box workers, especially 

e control dosimeter(s) were stored near these 

cations, this would have created additional exposures to the control dosimeter(s). In these 

dosimeter from the worker’s 

simeter reading to arrive at a net exposure for the worker would result in value less than the 

actual w ed

radiatio ental 

assessm

exposures in situations where the control dosimeters were stored in EALER locations. 

etry. 

diation fields at the RFP were perhaps not as complex as at some of the other 

cilities, they still present a challenge to dose reconstruction. One of the main concerns is the 

fact tha ve led

to neut the

fact. T

5.6.1 

Rocky Flats was a “first of a kind” facility relative to large-scale processing and fabrication of 

ssile materials for nuclear weapons. For this reason, quantifiable knowledge of neutron hazards 

e

the inherent limitations of personnel monitoring for neutrons prior to the development of more 

ea should be investigated to determine the extent of the possible underreported/miss

ecause of the practice of wearing the dosimeter under the lead apron. 

Location of Dosimeters When Working Around Gloveports/boxes 

Workers were generally instructed to wear their dosimeters at chest height. This combination 

created situations where a worker might receive doses to the head or lower parts of the bod

would not be registered by the dosimeter on the ch

the worker would receive unrecorded doses to the lower body. Also, a worker might

gl

badge worn on the chest. This issue was not identified or addressed in the TBD and sh

in

over extended periods. 

5.5.9 Elevated Ambient Levels of External Radiation Affect Net Dose 

Elevated ambient levels of external radiation (EALER) may have occurred at the RFP. This 

could have occurred at locations where radioactive materials were handled, transported, piped, 

stored, or dispersed into the environment. If th

lo

cases, the standard practice of subtracting the reading of the control

do

orker’s exposure being recorded, because the worker was exposed to this elevat

n field also. This missed dose would not be accounted for in the environm

ent procedure. Therefore, there is a need to account for EALER as occupational 

5.6 ISSUE 6: NEUTRON DOSIMETRY AND EXPOSURES 

Neutron dosimetry has historically been more difficult to perform than gamma/beta dosim

While the ra

fa

t neutron fields were not considered to be a problem in the early days; this could ha

ron exposures that were not monitored, and resulted in workers being badged after

his brings up the following areas of concern. 

The Likelihood of Unmeasured Neutron Exposures 

fi

in the workplace at Rocky Flats was acquired, on an iterative basis concurrent with large-scal

production, and from the experience and experimental evidence from other sites. In addition to 
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ess

g in and around different processing areas

(i.e., maintenance workers, pipefitters, sheetmetal workers, and electricians) were individually 

y

y Flats in 2001:

tetrafluoride compound.. Neutron

240

d primitive nature of

optimal measuring technologies in the early 1970s, it is not at all clear that employees in proc

areas, particularly those whose jobs required movin

monitored for neutron exposure. The lack of individual monitoring for neutrons is inferred b

the authors of a report summarizing data relative to the Former Radiation Worker Medical 

Surveillance Program at Rock

Significant neutron radiation exposure was possible from alpha/neutron reactions 

with light elements, especially from plutonium

exposure was possible from spontaneous fission of Pu during handling of large 

quantities of plutonium as metal, of 244

Cm, or of plutonium enriched in 240Pu for 

special projects conducted at the site. (Daugherty et al. 2001) 

Documentation exists which establishes the considerable uncertainties an

neutron measurements in the early years, as evidenced by the incomplete criticality safety 

program of the era, as outlined in Attachment 10. 

5.6.2 Use of Neutron Track Plates and Uncertainties in Dosimetry, 1951–1957 

The early history of neutron dosimetry at the RFP is illustrative of these uncertainties: 

1952–1956: The occupational external dosimetry TBD, (Langsted 2004, pp. 12 and 32), stat

that LANL processed neutron track “plates” for the RFP from 1951 to 1956 and that little was

known about their performance and calibration. Putzier (1982, pg. 43) mentions that it would

ore corre

es

be

ct to use the word “film” instead of the word “plates” when referring to the LASL 

3

as

en used a nnel neutron dosimeter.” 

o separate vendors at

m

service, but does not explicitly state that it was NTA film. In addition, a paper written in 196

by Mann and Boss (1963) states, “Since 1952, Eastman NTA (now called Type A) emulsion h

be s a perso

However, the NDRP (Falk et al. 2005, pg. 2) reports the following: 

The initial neutron sensitive element was known as a neutron track glass plate. 

These neutron track glass plates were pieces of glass approximately onesixteenth inch thick and one inch square with an emulsion coating on one side of 

the glass. 

And then on page 3, Falk et al. 2005 states that, “From 1952 through 1956, the neutron track 

plates were supplied, processed, and evaluated by LANL.” The NDRP report (Falk et al. 2005, 

pg. 3) goes on to confirm that the “plates” were not the same as NTA film by stating the 

following: 

NTA film was the neutron sensitive element used at Rocky Flats from 1957–1970. 

Rocky Flats procured their neutron-sensitive films from tw

different times based on availability and cost. These two vendors were either the 

DuPont Chemical Company or the Kodak Company. These films were created, 
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nd

hese conflicting statements from presumably reliable sources (some of which were present 

s at the

FP. It is difficult to ensure accurate claimant dose using the details of a given report, if the 

wrapped in a double-sided paper shield (white on the outside to reflect light a

black on the inside to absorb light), packaged in film boxes, and supplied to their 

customers. 

T

during the early years) create concern on the reliability of early neutron dosimetry record

R

historical system of dosimetry is not agreed upon. 

1956–1957: According to the TBD, (Langsted 2004, pp. 12 and 33), during 1956–1957 an 

outside source, by the name of HPS, performed the neutron dosimetry using NTA film, but lit

is known about this process. 

tle

1958-: The TBD (Langsted 2004, pg. 33) states that in 1958, personnel at the RFP took over th

responsibility of neutron dosimetry using NTA film. 

According to the LASL external dosimetry TBD (Widner 2005), LASL only used neutron plates

for neutron dosimetry during a short period from August 1949 to April 1951. This would

ive in the neutron-plate business during the time they were 

e

indicate that LASL was not act

In view of the uncertainties in the neutron dosimetry processes during the time period 1951 

through 1957, it is not apparent that the recorded neutron dose is correct or claimant favorable. 

More thorough investigation into neutron dosimetry during this period is needed to determine 

the det issed and 

undere ck plates

were t ressed

in the ium

worke he new

inform to

neutro

5.5.2.1

The us s to

neutro sed

neutro ). It

states d until

the ND om the TBD that this is an 

lutonium workers. Therefore, a 

more thorough and detailed analysis of missed or underestimated neutron dose needs to be 

perform NDRP

report rt is

supplying and processing neutron plates for the RFP during 1952–1956. 

ails of the process. This would enable the DR to more accurately correct for m

stimated/unrecorded neutron doses. It should also be determined if neutron tra

he same as NTA film, and if not, their characteristics. Many of these issues are add

NDRP report (Falk et al. 2005) issued February 7, 2005, but are limited to pluton

rs only. The neutron dose sections in the TBD need to be revised in view of t

ation in the NDRP, and extended to include other, non-plutonium workers exposed

ns.

Use of Neutron/Photon Rations to Generate Correction Factors Needs More 

Review

e of neutron/photon ratios and comparing valid overlapping NTA film reading

n TLD results, in order to generate correction factors for under-monitored or mis

n doses, was briefly addressed in the TBD (Langsted 2004, Section 6.7.3.4, pg. 35

that the two correction factors (1.99x and 1.13x) provided in Table 6-19 will be use

RP report (Falk et al. 2005) becomes available. It appears fr

interim position on neutron dose, and that the final results will be based on the NDRP report.

However, the NDRP report only addresses neutron doses for p

ed for non-plutonium workers that were exposed to neutrons. The results of the

cannot be directly applied to other neutron exposures at the RFP, because this repo
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especia tors

(CF) sp

other neutron spectra encountered by other workers at the RFP. 

itially, the n/p

ined to be around 1:3 in the final-assembly area, but was later found to be 3:1 in the 

ilar analysis is needed to include other non-plutonium workers 

ed

concern that neutron dosimetry at the RFP, especially in the early 

e time, it was 

ed that the energy level was between 0.25 and 0.4 MeV. They subsequently realized that 

ype A emulsion is between 0.25 Mev(1)

 and 

(2)

measurements of neutrons below about 0.75 Mev in energy. 

lly formulated for a few plutonium neutron spectrum situations with calibration fac

ecific for those energy spectra. Therefore, these spectra would not necessarily match 

In addition, the neutron/photon values changed with location and with time. In

was determ

chemical production area. Recent data showed some areas had an n/p value of 12:1 

(Bistline 2005). As stated above, the neutron dose sections in the TBD need to be revised in 

view of the NDRP report, and sim

exposed to neutrons. 

5.6.3 NTA Neutron Energy Threshold May Lead to Missed Dose in RFP Records 

NTA film has a 0.8 MeV to 1.0 MeV threshold for neutron-proton recoil created tracks read on a

routine, high through-put, basis; even under controlled conditions with good quality control. 

The shortcomings of past dosimetry methods using NTA film and assuming a neutron energy 

threshold of 500 keV, or even less (Wilson et al. 1990), has been recognized and has caus

some facilities to reevaluate their neutron dosimetry records. 

In view of this fact, there is 

days, could have missed significant doses of lower energy neutrons, because at th

assum

this energy threshold for neutron detection might be too low and set a sensitivity at 0.48 – 

0.75 MeV. Later research shows that this neutron energy threshold was still too low. 

A paper written in 1963 by Mann and Boss of the RFP (Mann and Boss 1963) states the 

following: 

I. Introduction 

Since 1952, Eastman NTA (now called Type A) emulsion has been used as a 

personnel neutron dosimeter. Early calculations of neutron flux from certain 

plutonium processing vessels and tanks indicated that the film was not seeing all 

of the fast neutrons present in a given area. Even when the neutron calibration 

source was changed from PoBe to PuF4, there was still a large discrepancy 

between film, calculated, and survey meter dose. Moderated B1

° or B'3 detectors 

in one form or another have been used for survey work. 

The effective neutron energy cutoff of T

0.4 Mev , depending on the track-reading system used and the ability of the 

reader. The sensitivity of the film thus becomes quite low for neutrons whose 

average energy is in this range. The sensitivity is 20 mrem/track/10 mm2

 for 

0.48-Mev (average) neutrons compared to 6.8 mrem/track/10 mm2

 for 1.4-Mev 

(average) neutrons from PuF4. This same situation exists for most proton-recoil 

chamber instruments, making the use of either device extremely difficult for dose 
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ffered the same shortcomings as NTA film. Then at the end 

f the paper, Mann and Boss go on to conclude the following: 

Type A film can be used as a personnel neutron dosimeter provided calibrations 

r

llows. 

NDRP Does Not Cover Non-Plutonium Workers 

ome of those neutron dosimetry issues noted above are addressed in the Neutron Dose 

film

. Some

d other sources of neutrons such as (alpha, neutron) 

actions in UF4, criticality experiments, calibration sources, etc., besides Pu at the RFP during 

this

5.6.4.2 Use

The use of n/p TLD results

in order to gen efly

addressed in th ction

factors (1.99x et al.

2005) become position on neutron

dose, and that eport

only addresses etailed

analysis of mi nium

workers that w directly

applied to othe few

plutonium neu ould not

nec i

abo

similar an

Not only was the NTA film missing doses resulting from lower energy neutrons, but also the

proton-recoil survey instruments su

o

are made with neutron sources having average energies the same as those being 

measured. Larger field areas must be read for lower energy neutrons to minimize 

the statistical errors and to reduce the minimum detectable dose. (Mann and Boss 

1963) 

5.6.4 NDRP Report Lacks Important Dose Reconstruction Information 

The NDRP report (Falk et al. 2005) is a very detailed analysis of neutron dose reconstruction fo

plutonium workers at the RFP. However, there are limitations to applying the report as fo

5.6.4.1 

S

Reconstruction Protocol (Falk et al. 2005) that covers the period of 1952–1970 when NTA

was used for neutron dosimetry. However, it addresses only workers in the plutonium facilities; 

it does not cover workers who could have been exposed to other sources of neutron radiation, 

especially during the early days before some of the sources of neutrons were recognized

workers were exposed to uranium an

re

time period when NTA film procedures are in doubt. 

of Neutron/Photon Ratios and Film/TLD Comparisons Needed 

ratios and comparing valid overlapping NTA film readings to neutron

erate correction factors for under-monitored or missed neutron doses was bri

e TBD (Langsted 2004, Section 6.7.3.4, pg. 35). It states that the two corre

and 1.13x) provided in Table 6-19 will be used until the NDRP report (Falk

s available. It appears from the TBD that this is an interim

the final results will be based on the NDRP report. However, the NDRP r

neutron doses for plutonium workers. Therefore, a more thorough and d

ssed or underestimated neutron dose needs to be performed for non-pluto

ere exposed to neutrons. The results of the NDRP report can not be

r neutron exposures at the RFP, because it is especially formulated for a

tron spectrum situations with specific CF for those energy spectra and w

essar ly match other neutron spectra encountered by other workers at the RFP. As stated 

ve, the neutron dose sections in the TBD need to be revised in view of the NDRP report, and 

alysis is needed to include other non-plutonium workers exposed to neutrons. 
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5.6.5 N

Page 34,

to range f

value with a resulting multiplying factor of 1.79. It would appear that the higher value of 60% 

tron

his

of

e

arge

ents addressed. These issues are discussed in the subsequent 

sections. 

.7.1 Unmonitored Individuals 

here is concern that some workers, especially during the early years, may not have been 

The

DRP report (Falk et al. 2005), page 1, states the following: 

kers, neutron monitoring was not provided until

the early 1960s, and their dose of record may not include significant 

und that neutron

shielding was present for these operations until the early to mid 1960s. 

eutron Dose Multiplication Factor

Table 6-18, lists the potential missed neutron dose below 800 keV for early NTA film

rom 16% to 60%. However, the text below it selects 56% as the claimant-favorable 

missed neutron dose with a resulting multiplying factor of 2.5 should have been selected (i.e., 

1/1-0.60) = x2.50). In addition, even if the value of 56% was selected for the percent of neu

dose missed, then the multiplying factor would be 1/(1-0.56) = x2.27 instead of 1.79. T

information needs to be corrected. 

5.7 OTHER EXTERNAL DOSIMETRY ISSUES 

A combination of other factors have been found that SC&A feels complicates and makes 

difficult the ability to fully characterize the external dose to workers. This may be due to the 

lack of issued dosimetry in the early days, or missing or incomplete dosimetry. Some aspects

extremity badging may prevent full characterization of extremity dose. The VARSKIN softwar

may miss some of the important radionuclides contributing to skin dose. There is little 

information in the TBD (Falk 2004) that discusses the possible use of industrial x-ray units, l

radioisotope sources, or neutron generators that SC&A feels could be potentially significant 

contributors to external dose. Neither is the subject of past or recent D&D activities and their 

unique monitoring requirem

5

T

recognized as having the potential of receiving radiation doses from their work at the RFP.

occupational external dosimetry TBD (Langsted 2004, Section 6.9.1, pg. 42), states that in the 

early 1950s, only groups expected to received doses greater than 10% of the RPG were

monitored. However, we find that: 

(a) The N

For some plutonium wor

contributions from neutron exposure received prior to being issued a 

neutron dosimeter. These workers included most of the employees working 

in Building 71 (now Building 771). Only a small number (10–18) of these 

employees were monitored for neutron exposure, and that monitoring was 

only during the period October 1956 to September 1957. 

Operations in Building 71 involved chemical processing of plutonium in 

acid solutions and resulted in significant neutron fields from the alphaneutron reaction with light elements, especially from the plutonium 

tetrafluoride compound. No evidence has been fo
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(b) Column 2 of Table 2, pages 17 and 18, of the RFP annual radiation exposure report for 

ment entitled Film Procedure Timeline for the Neutron 

Dose Reconstruction Project 2003 (RFP 2003, pg. 6), provides the number of NTA 

f g the two tables for the years 1959–1969 shows 

that for the total number of workers badged, on the average, only approximately 25% 

nd

n administrative decisions, not necessarily on actual, well conducted, survey results (if this was 

n as in areas as stated in the 

DRP report (Falk et al. 2005) and as indicated in some former worker interview sessions). The 

azards

sociated with the operations at the RFP (Putzier 1982). This was coupled with the need to 

sometim iews).

This sit s and

categor eal

possibi DR by using co-worker dose or area monitoring 

formation, because these potential areas of exposure were not deemed necessary to monitor at 

Photons with E > 250 keV 

re

could have an affect on the final

assigned dose. 

5.7.2.2 Use of VARSKIN Software 

Page 42, the last paragraph of Section 6.8.6 recommends using the VARSKIN software to 

calculate skin dose from contamination. However, the TBD (Langsted 2004, Table 6-26, pg. 43) 

shows that the VARSKIN Mod 2 only includes one (234

Th at 3.40E-07 Ci/gm of DU) of the two 

the year 1984 (RFP 1985) provides the total badged personnel per year for the years 

1953–1984. Another RFP docu

ilms processed per month. Analyzin

were issued NTA film badges (over the 11-year period, it ranged from 9% to 50%). 

For example, in 1960, a total of 1362 workers were badged throughout the year a

approximately 300 NTA films were processed per month (therefore, 300 workers were 

badged for neutrons throughout that year), this results in a ratio of 300/1362 = 22%.

These facts bring up the question of if the policies and procedures in place during this period at 

the RFP were adequate to ensure that the workers who were at risk of receiving significant 

radiation doses (especially neutron doses) were actually, and adequately, badged. The 

assignment of dosimeters to workers using the “10% of the RPG” policy was apparently based 

o

ot the c e, then there would not have been lack of monitoring in certa

N

basic problem with external radiation monitoring at the RFP for 1951 through the early 1960s

was the beginning philosophy that there would not be a great deal of external radiation h

as

es put production ahead of other issues (as stated during former workers’ interv

uation led to some after-the-fact badging and catch-up monitoring for some area

ies of workers. Missed doses caused from these practices were therefore a very r

lity and difficult to compensate for in 

in

the time. 

5.7.2 Photon and Beta Dose Determination 

5.7.2.1 

Photon exposure spectra are assumed to be mostly in the 30-250 keV range to be claimant 

favorable (Langsted 2004, Section 6.6.1.1, pg. 24). However, if there are any DR cases whe

shielding (either external or internal to the body) must be considered, then photons with energies

>250 keV may contribute more to the final effective dose because of their greater penetrating 

power. It needs to be determined if there are cases where this
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m tiv g 234mPa at 3.40E-07 Ci/gm of DU (the third isotope, 

U at 3.40E-07 Ci/gm of DU is not itself an active beta emitter)]. Page 37 states, “Thus, for 

,

this section:

1

ture. It

o did wear wrist badges were not monitored for extremity neutrons, 

t only for beta/photon. For those handling neutron-emitting materials, the neutron doses to the 

extrem ly issued

film ba t recorded. 

5.7.3.2 

In some ned as the extremity dose of record. 

This is not claimant favorable. The TBD (Langsted 2004, pg. 43) states the following: 

id not receive extremity (wrist) dosimeters. In such cases, 

the wrist (forearm) dose was assigned as the measured skin (shallow) dose [as 

e

NDRP esponse as

only 40% of the wrist photon dose. Additionally, in most cases the dose to the hands would be 

s

measured by the WB badge. 

The pre forearm,

and cer t

dosime sible

extrem information. 

In these uld be greater

than th

ost ac e beta isotopes [the other bein

238

depleted uranium, one is dealing essentially with 2.29-MeV (Emax) beta particle from 234mPa,

the most energetic contributor to the beta exposure.” With the information provided in the TBD

it would appear that VARSKIN should not be used in DR if it leaves out 234mPa. Including 

234mPa in the DR process is necessary to arrive at correct doses. 

5.7.3 Extremity Dose 

The TBD addresses extremity dosimetry issues on page 43, Section 6.10. However, there are 

several items of concern in

5.7.3. Use of NTA Film in Wrist Badges 

Generally, neutron doses to the extremities are not a large issue, but at the RFP (more so than at

most facilities) there appears to have been considerable hands-on work with uranium and 

plutonium compounds that had the potential to deliver neutron doses to the extremities. Between

1951 and 1970, beta/gamma film was used for extremity dosimetry. However, no mention of 

NTA film being used for extremity monitoring was mentioned in the TBD or other litera

appears that the workers wh

bu

ities could have been greater than the whole-body dose monitored by the regular

dges. Therefore, this is a potential source of worker’s missed dose that was no

Extremity Dose Assumed to be Equal to Whole-Body Dose 

 cases, the measured whole-body dose was assig

Many RFP workers d

measured by the whole-body badge], and the hand dose was assigned the sam

value.” (Langsted 2004) 

(Falk et al. 2005, pg. 22) lists the average whole-body (WB) photon dose r

greater than the dose to the wrist by 2.5x to 5x (Mann 1964). This would mean that the hand

could have received 6x to 12x the dose as

vious procedures at the RFP do not appear to be claimant favorable because the

tainly the hand, could have received a higher dose than the whole body. If wris

ters were not used, the DR should consider the worker’s job description for pos

ity exposures, and calculate an appropriate dose from the whole-body dose

cases, the wrist dose would be greater than the WB and the hand dose wo

e wrist dose, as per the applicable hand-to-wrist ratio. 
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5.7.3.3 

It appe seriously

ffect claimant dose reconstruction at the RFP. For example, Section 6.10, page 43, states that 

ons.

n the TBD. The ratios need to be developed and 

sed in the DR, if needed.

.7.4 Industrial X-ray Units and Neutron Generators 

non-destructive testing on the parts after they were finished. They conducted 

a neutron generator. They also conducted x-ray testing to look at welds and 

oximately 10 employees

worked in this area of 444. [Emphases added] 

After the parts were completely assembled, they would undergo a number of 

testing operations to determine their suitability. Parts would be x-rayed in 

Module E to identify any structural flaws. A number of chemicals were used in 

the development of the x-rays which was done on site. [Emphases added] 

And in Industrial Hygiene Evaluation of Building 771 Exposures (Ruttenber 1996, pg. 35), 

Ruttenber refers to an X-ray Laboratory: 

The X-ray Laboratory was used to determine the amount of plutonium in solid 

samples received from Production. This laboratory was used when plutonium 

concentrations were expected to be more than one g/l. Magnesium oxide and 

ammonium hydroxide were used to suspend the solution so the plutonium 

concentration could be measured by X-ray fluorescence. The matrix for 

suspension was mixed every three to six months. Periodically, the odor of

Valid Hand-to-Wrist Ratios 

ars that there is a lack of valid information on hand-to-wrist ratios that could

a

the details on the hand-to-wrist ratios are not available. The last sentence in this section states 

that additional information on the hand-to-wrist ratios is required before shallow doses to the 

extremities can be reconstructed. The last paragraph on page 37 states that surface beta dose 

rates on the order of 1 to 20 rad per hour had been observed, and RFP had problems with 

elevated beta dose rates from contamination on leather gloves worn during foundry operati

No values for hand-to-wrist ratios are provided i

u

5

The TBDs (Little and Meyer 2004; Flack and Meyer 2004; Langsted 2004) do not mention any 

use of industrial x-ray units, large radioisotope sources, or neutron generators for R&D or nondestructive testing (NDT), or any associated radiation exposure or monitoring. However, these 

units (some in the MeV energy range) were apparently used at the RFP at sometime as referred

to in an article by Ruttenber, Industrial Hygiene Evaluation of Building 444 Exposures

(Ruttenber 1996, pg. 17): 

The Radiography Laboratory was staffed by NDT Technicians who performed 

density testing using Freon and determined the oxygen content of beryllium using 

performed die penetration testing on the parts. Appr

Additionally, in Industrial Hygiene Evaluation of Building 707 Exposures (Ruttenber 1996, 

pg. 4), Ruttenber refers to: 
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ammonium was present in the air at the beginning of the shift. Prior to the 

constructi re in

 area. [Emphases added]

In view ntiall ates that accompany these units, and the fact 

that the of dosi s m

neutron at of th r plutonium and uranium isotopes, this is an area 

that needs to be addressed. The number of units, energies, periods of operations, operating 

procedu radiation exposures, and if 

radiation doses were under-recorded or missed. 

5.7.5 Dec

Deconta nd de issioning (D&D) ac sent no s and 

unique e inte ring requireme idering any years of operations 

involving unique materials at th FP, D&D activ both t ve years of the plant

and during recent clean-up activities present radio itions lly encountered.

Some of these unique conditions include high-fired Pu and Am in ductwork and in building 

structures, and also a variety of ioisotopes in the soil (some that h een identified are 

89/90Sr, 1 238Pu208 212Pb, 214

Pb p). If th r radiological

hazards with i were not recognized before D&D began, under-recorded 

and/or m ose is pr ese issues were ioned or the TBD. 

5.8 ISSUE 8: HISTORIC OPERATIONS AND RADIOLOG

5.8.1 te listi ides of C

The TB scripti ovides useful information regarding radionuclides handled in 

different buildings, but it does not list the general and quantitie

workers been ed. As Table 11 ind as developed for off-site 

dose reconstruction studies (ChemRisk 1992). 

on of Building 559, the X-ray and Radio Assay laboratories we

separate rooms; now they are located in the same

of the pote y high radiation exposure r

response metry yste s may be different for these units (high-energy photons and 

s) from th e standard calibration fo

res, etc., needs to be determined to assess the potential

ontamination and Decommissioning Activities Not Addressed 

mination a comm tivities pre nmroutine situation

xternal and rnal monito

e R

nts. Cons the

ities during he acti

logical cond

241

not norma

rad ave b

34/137Cs, U,

d

, Pu, Tl, 

D activ ties

239 , and 237N ese, and othe

, associate D&

issed d obable. Th not ment addressed in

ICAL CONTROLS 

Inadequa ng of Radionucl oncern

D Site De on pr

types s of radionuclides to which 

may have expos icates, such a listing w
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Table 11. Radioactive Materials at the Rocky Flats Plant 

Radioactive Material Handled in Kilogram Quantities 

Americium 241Am

Plutonium 238Pu, 239Pu, 240Pu, 241Pu, 242Pu

Thorium 232Th

Uranium 233U, 234U, 235U, 236U*, 238U

Radioactive Materials Handled in Gram Quantities (<1 Kg) 244Cm 3

H (Tritium) 237Np 228Th 

Other Sources – 

Includes sealed solid sources, plated sources, liquid sources and analytical stock solutions 

Actinium 228Ac Nickel 63Ni 

Aluminum 26Al Plutonium 236Pu, 244Pu

Americium 243Am Polonium 210Po 

Antimony 124Sb, 125Sb Potassium 40K 

Argon 39Ar Promethium 147Pm

Barium 133Ba Protactinium 231Pa, 234Pa 

Beryllium 7Be Radium 226Ra 

Bismuth 207Bi, 210Bi Ruthenium 106Ru

Cadmium 109Cd Selenium 75Se 

Californium 250Cf, 251Cf, 252Cf Silver 110Ag, 110mAg 

Carbon 14C Sodium 22Na 

Cerium 139Ce, 144Ce Strontium 85Sr. 89Sr, 90Sr

Cesium 134Cs, 137Cs Technetium 99Tc, 99mTc 

Chlorine 36Cl Thallium 204Tl 

Cobalt 57Co, 60Co Thorium 230Th, 231Th, 234Th 

Curium 245Cm, 246Cm Tin 113Sn 

Europium 152Eu, 154Eu, 155Eu Uranium 232U, 236U

Holmium 166mHo Ytterbium 169Yb 

Iridium Ir Yttrium 88Y, 90Y 192

Iron 55Fe Zinc 65Zn 

Krypton 85Kr 

Manganese Mn 54

Mercury 203Hg 

Source: ChemRisk 1992, Appendix A (DOE 2000)

5.8.2 The Need for Incident Database Access for Dose Reconstruction 

in TBD approach should be augmented with the 

clusion of relevant databases of incidents. NIOSH should not rely on the a priori assumption 

the

The TBD provides salient details relative to illustrative incidents, which are the most extensive 

s reviewed to date by SC&A. This superior

in

that the files of individual claimants contain all relevant information on incidents involving

claimant. For monitored workers, ORAU and NIOSH rely primarily on the worker dose record 

and the Computer-Assisted Telephone Interview (CATI) for potential exposure data on

radiological incidents. Workers have often complained of poor record keeping and fabrication of

records, which have been confirmed in congressional and Tiger Team investigations. The Rocky 

Flats Site Profile, as it now stands, does not have a dedicated database or assessment for off-
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al doses) or record keeping practices that might indicate problems with the 

tegrity of individual worker data in some settings and periods. Such data bases exist for Rocky 

Fla

public d

in

This co

collecte

which a

(Daugh

2,000 rem) TEDE, with about 10% exceeding 1 Sv (100 rem)...for 

those participants with TEDEs above 1 Sv (100 rem), the majority of the dose 

5.8.3 

The po r

materia and

econtamination and decommissioning. All of these activities have resulted in employees 

wor lative

to radia

5.8.4 ence of

By 9 in

thou an

plutoni

conting

rge inventories of excess nuclear materials, stored under sometimes questionable circumstances 

from a radiological standpoint. Moreover, post-production fissile material storage problems

normal exposures and incidents that can guide dose reconstructors as to missed incidents (and 

hence addition

in

ts which should be incorporated into dose reconstruction efforts. According to the 1992 

ose reconstruction study done for the State of Colorado:

The most complete historical record available of all accidents at Rocky Flats is 

maintained by the Occurrence Management Department of EG&G Rocky Flats

the form of Summary of Events (SOE) database that covers the period from 1952–

1990...At the time of ChemRisks’s Review, the database contained approximately 

1,767 accident entries. (ChemRisk 1992, pg. 211) 

ncern, particularly over significant exposures due to incidents, is underscored by data 

d by the Former Radiation Worker Medical Surveillance Program at Rocky Flats, in 

pproximately 825 former workers participated from 1992 through December 1999. 

erty et al. 2001) According to an analysis of this program:

Doses for program participants seen during this period range from near zero to 

nearly 20 Sv (

typically is from internal depositions of plutonium...Approximately 10% of the 

participants seen during this period have lung doses of 6 Sv (600 rem or more 

and the maximum is about 120 Sv (12,000 rem)...Approximately 10% of those 

bone surface doses are 5 Sv (500 rem) or more, and the maximum is about 

320 Sv (32,000 rem )...The causes of many of the higher internal doses were 

accidents that generally are well documented... [Emphasis added.]

(Daugherty 2001) 

Inadequate Description of Post-Production Mission at Rocky Flats 

st-production mission, now in its 16th year, has involved nuclear material storage, nuclea

l stabilization and packaging, environmental restoration, waste management,

d

king in various radiation environments, which should be described and characterized re

tion exposures and dose reconstruction. 

Inadequate Description of Nuclear Plutonium Storage Activities and the Abs

Related Dose Reconstruction Guidance 

19 4, 2 years after production ceased, the RFP was storing metric tons of plutonium

s ds of individual packages—a considerable portion of the entire DOE inventory of excess

um at the time. Like other DOE facilities, the RFP had not established storage

encies to address the end of production (DOE 1994, pg. 18). This resulted in sudden and 

la
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pos a

For ins

• ajor 

cording to a 1994 environmental safety and 

health vulnerability assessment performed by the Energy Department, neutron gamma 

e from <1 to 10. 

Previously the neutron/gamma ratios could be characterized by location due to the type of 

om plutonium

met ium

solu future

wor cally” 

[Emphasis added] (DOE 1994, pg. 105). This matter is of concern, because the external 

ty Board

(DN plied in the

cas n dose

calc

• The TBD does not address the potential for missed doses due to radiological control 

ing in several areas of the 

building; Solutions in tanks, bottles, and piping ...[and] a large number of waste drums in 

ew of

...a

”

e

compliance

d

e Rocky 

• m did not appear to be fully developed. RFP 

personnel noted during briefings that the Radiation Work Permits (RWP) do not stipulate 

nel

indicated that contractors may come and go and not be bioassayed. A deliberate 

• 

lities

the

ed much different set of dose reconstruction challenges than encountered during production.

tance: 

NIOSH has not addressed nor provided dose reconstruction guidance relative to the m

shift in neutron gamma ratios, associated with large-scale fissile material storage 

problems during the post-production. Ac

ratios changed “dramatically and [have] become more variable with the recent 

consolidation of material...[and]... now typically range anywher

plutonium operation and material handled. Historically, the typical dose fr

al was 30 percent neutron, 70 percent gamma, while that from liquid pluton

tions was 70 percent neutrons, 30 percent gamma. Of particular note for

k is building 317, where the thermal neutron flux has increased dramati

dosimetry program algorithm was found by the Defense Nuclear Facilities Safe

FSB) in 1993 to have “a wrong correction factor [that] may have been ap

e of the K17 dosimeter chip,” potentially impacting the accuracy of neutro

ulations.

problems, which resulted in “gross contamination [and] contamination from inactive 

gloveboxes and tanks; leakage of rainwater through the ceil

operating areas” (DOE 1994, pg. 70). This concern is underscored by a 1993 revi

occupational radiation protection at Rocky Flats by the DNFSB, which reported: “

potential existed for workers to be exposed to radiation without being monitored...

Subcontractors working at the plant did not appear to be totally integrated into som

Radiation Protection Programs. Currently, their contracts do not require

with the RCM (DOE’s Radiation Control Manual), which requires consistent an

uniform dosimetry for workers in radiation areas (DNFSB 1993). (At that tim

Flats was in compliance with only 22.4% of the requirements in the RCM.) 

“Management of the bioassay progra

the requirements for bioassay...when asked about other subcontractors, RFP person

delinquency tracking system does not exist” (DNFSB 1993). 

“Air monitoring equipment in the workplace at [the Rocky Flats Plant] is not in 

compliance with the requirements of the RCM.” Continuous air monitoring capabi

were found be approximately 7 times less efficient and accurate than required by

DOE (DNFSB 1993).
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l

w

the assessment:

rned about the 

potential for plutonium fires from the vibration or movement of packages. 

nt

.

om,

radioactive material content. (DOE 1994, 

pg. 77). 

The nat d

in the May 1995 DNFSB Staff report, which noted the following: 

f

inated organic compounds that could result in violent reactions, 

The risks of the post-production m

tible material, a fire occurred

vebox. A significant firefighting effort

• Documented respiratory protection was not in compliance with DOE’s Radiation Contro

Manual (DNFSB 1993). This raises important questions about potential missed doses, 

which NIOSH should address, particularly since in 1994 it was reported that: “ many 

rooms require full face masks due to airborne contamination or undesirable air flo

within the room. The former condition is caused by surface contamination becoming 

loose in the room” (DOE 1994, pg. 38). 

Of those, 81 vulnerabilities were identified by the DOE in 1994 as having potential adverse

impacts on workers (DOE 1994, pg. 18). According to

• In touring the facilities, several people were very conce

However, the concern about the perils associated with package moveme

restricts the ability to inspect the packages adequately (DOE 1994, pg. 26)

• Building 776 also contained.... ...near the advanced size reduction ro

mixed waste drums are stored. Although all of the drums had labels and

travelers and other documents included estimated contents, a number of 

drums have not been assayed for

ure and extent of work necessary to carry out the post-production mission is exemplifie

EG&G stated that there are 17.5 metric tons of combustible residues containing 

about 0.5 metric tons of plutonium. The total plutonium inventory of all residues 

at Rocky Flats is 3.1 metric tons. The combustible residues consist primarily o

paper, cloth, filters, resins, wood and various plastics along with small amounts 

of oils, greases and solvent. Potential hazards in the combustibles include 

radiolytic generation of hydrogen and other flammable gases, nitrated organic 

compounds which could be flammable or shock sensitive, plutonium metal in 

contact with chlor

packaging degradation due to chemical and/or radiolytic effects, and radiation

exposure to workers (DOE 1995). 

ission at Rocky Flats are underscored by the following: 

On May 6, 2003, due to the accumulation of combus

in the basement of Building 371 at the Rocky Flats Environmental Technology 

Site (RFETS) as workers were preparing to remove Glovebox 8 from the facility. 

The fire broke out after operators began cutting a hole near the top of Glovebox 8 

to establish a ventilation path for the glo

ensued, including the discharge of more than a dozen fire extinguishers and 

eventual use of a fire hose (DOE 2003). 
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As deta

require curate or

mis g the

com e

accordi

expand

accoun nd the strong potential for missed and/or 

inaccur

rad o

lated exposures to radiological and non-radiological hazards done at the 

knowledge

ive ecific activities

otentially hazardous areas

al incidents, 

ties (Alvarez 2005).

 in production areas on a daily basis 

ork maintenance and repairs. Pipefitters worked 

tal workers worked on ventilation lines. In 

ired to enter gloveboxes to perform repairs 

• Maintenance Machinists spent a large portion of their time repairing pumps in 

ere

lls

iled above, the post-production period of the past 18 years created circumstances which 

d important adjustments to occupational radiation dosimetry, so as to avoid inac

sin external and internal doses. Some of these adjustments may hold implications for

pl teness and accuracy of ongoing and future dose reconstructions. NIOSH should 

ngly expand the current site profile scope to include the post-production period. This 

ed scope should address the preceding issues and provide guidance which takes into 

t the impact of stored excess fissile materials a

ate doses due to the documented problems in the overall quality of the Rocky Flats 

iati n program shortly after the end of production.

5.8.5 The Need to Utilize Job Exposure Matrix Information 

The TBD acknowledges that work-related job exposure analyses performed at the 

University of Colorado would add further insights to the dose reconstruction work at 

Rocky Flats. The SC&A team was allowed a limited review of working papers relative 

to job-re

University of Colorado. We find that these analyses provide important

relat to the history and nature of building processes, job titles, and sp

performed. For instance, in Building 771: 

• Radiation exposures to production workers (e.g., Chemical Operators, Process Operators, 

and Assistant Chemical Operators) influenced the rotation of workers through process 

areas (Alvarez 2005). 

• Radiation Protection Technicians were frequently present in p

because they were involved in the discovery of and/or response to radiologic

as well as subsequent repair activi

• Pipefitters and Sheetmetal workers were

performing piping, glovebox, and duct-w

usually

on lines carrying liquids in areas and Sheet me

some instances, Sheetmetal workers were requ

(Alvarez 2005). 

gloveboxes and, in some instances, were required to enter the gloveboxes. Maintenance 

Machinists were also likely to encounter spills during repairs (Alvarez 2005). 

• Electricians worked along with Pipefitters, Sheetmetal Workers and Machinists and w

likely to receive similar exposures (Alvarez 2005). 

• Painters were often present in radiological environments, because a great deal of their 

time involved scraping contaminated surfaces and decontamination painting of spi

(Alvarez 2005). 

• Prior to 1972, filter changeouts involving potential exposure to radionuclides were 

performed by Production Workers and Sheetmetal Workers. This involved changing, 
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nd in

oted previously, NIOSH recognizes the importance and relevance of 

ational Environmental Dose TBD, ORAUT-TKBS-0011-3 (McDowell-Boyer and 

ittle 2 to perform

occupa l TBD

had sev

or source term and exposure pathways re-examination 

or a timeline for phases of operations, and data availability and types 

(3) 

(4) 

(5)

These i following sections.

5.9.1

The TB releases 

om 19 to environmental doses from

The TBD also excludes the impact of resuspension of soil at areas of the site other 

.

testing, and packaging filters designed to trap radionuclides in plenums and gloveboxes.

After 1972, Filter Technician became a separate job category (Alvarez 2005). 

• Very little work done by carpenters was done in production areas (Alvarez 2005). 

• Janitor, Handyman, and Utility Workers in many instances were not usually fou

production areas because most were awaiting clearances. However, some remained in 

these jobs for several years (Alvarez 2005). 

• Work done by Laboratory Workers was not in production areas, but involved glovebox 

work with smaller quantities plutonium (Alvarez 2005).

This job exposure matrix is important for the dose reconstruction process because it provides a 

basis for using co-worker data for filling in missing data for unmonitored or inadequately 

monitored workers. As n

this matrix, and has made a concerted effort to obtain it, but to no avail as of the date of this 

review. 

5.9 ISSUE 9: OCCUPATIONAL ENVIRONMENTAL DOSE 

The Occup

L 004) was evaluated to determine if adequate guidance was provided

tional environmental dose reconstruction. SC&A determined that the environmenta

eral deficiencies, including the following: 

(1) Need f

(2) Need f

Data adequacy and completeness 

Ambiguous recommendations for particle size 

Uncertainty with the RATCHET model 

ssues are discussed in the

Soil Resuspension 

D provides the annual inhalation intake for 239+240Pu and 241Am for atmospheric

fr 53–2002. It acknowledges that workers were subject

incidents, such as venting of material outside buildings or resuspension of contaminated soil 

(McDowell-Boyer and Little 2004, pg. 13). The internal dose from resuspension of 

contaminated soil has not been fully addressed by the TBD. The dose consequences from

contaminated soil seem to be limited to external dose. As indicated for the Savannah River Site 

reviews, there are a couple of methods to estimate inhalation exposure from resuspended 

radionuclides. These methods include the dust-loading approach and the resuspension-factor 

approach.

than the 903 Drum Storage Pad, which may or may not be represented by the air sampling data
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at the

ate does not include releases due to

suspension of contaminated soil downwind of the 903 Drum Storage Pad. On the same page, 

the TBD was clear in declaring that there were 24 identified discrete events during the period 

between 1964 and 1969, and the asphalt pad placed over the 903 Area in 1970, did not stop the 

resuspension of plutonium-contaminated soil to the air in later years. Releases from the 903 

Dru ated 3.1 Ci over several years (McDowell-Boyer and Little 

2004, pg. 9). Resuspension of soil from this and other surfaces during high winds or soil 

rem lt in both inhalation and ingestion of radioactive material, and 

sho

ith respect to external dose, the TBD focused on the incidental releases at the 903 Area Drum

active waste sites are one of the principal causes of groundwater 

urther

1989, the Environmental Protection Agency and the Federal Bureau of Investigation (FBI) 

initiated an investigation and regulatory, as well as legal actions, in response to clear 

SC&A has not seen any evidence that the resuspension of plutonium and americium was taken

into consideration in determining the internal dose values for the pre-1965 period. The 

occupational environmental dose TBD (McDowell-Boyer and Little 2004, pg. 9) stated th

total routine (non-discrete-event) plutonium emissions from 1953 to 1989 are estimated to be on

the order of 0.12 Ci (Voillequé 1999c). This estim

re

m Storage Pad totaled an estim

ediation activities could resu

uld not be excluded.

W

Pad, but did not include other soil contamination areas in its discussion. During a Government

Accountability Office (GAO) evaluation of the major problems at RFP (GAO 1988), 

environmental contamination at RFP was defined as a significant problem. The GAO stated in 

its report: 

In

contamination. A number of such sites were identified at Rocky Flats as potential

sources of groundwater and soil contamination. 

F more, 

A total of 108 inactive sites have been identified at RFP. Some of these sites are 

considered to be existing or possible sources of significant environmental 

contamination. Furthermore, some off-site areas have been contaminated with 

low-levels of plutonium. 

Of the 108 inactive waste sites identified, RFP officials have given a high priority 

to 27 sites, which have been grouped together in four areas because of their 

general proximity. These areas are (1) the 881 Hillside area, (2) the 903 Pad 

area, (3) the Mound area, and (4) the East trenches. 

Appendix C of the RFP’s Final Buffer Zone Sampling and Analysis Plan indicates that some

samples at the periphery of the site variously contained 2,830 pCi/g 137Cs; 2,610 pCi/g 238Pu; 

3,246 pCi/g 239Pu; and 149,060 pCi/g 208Tl. These obviously elevated activity levels in surface 

soils raise a concern over potential worker contact with radioactively contaminated soils. Some

past practices have been identified (and described in Appendix C) that would also add to these 

concerns. 

In
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),

WA), and the National Pollutant Discharges Elimination System

PDES) program (Lipsky 1989). During the investigation, the following areas were identified 

th mixed waste. 

• 

• 

• 

• 

• 

effluents from the Sewage Treatment Plant (STP) were discharged 

areas, NIOSH should review soil contamination data for the 

ation activities and the potential for exposure from either the areas 

noff. The high-wind events at RFP can also lead to contamination 

environmental law violations at RFP. Regulatory non-compliance was established under the 

various environmental statutes, including the Resource Conservation and Recovery Act (RCRA

the Clean Water Act (C

(N

as outside areas contaminated wi

• 750 Pad 

• B-series ponds 

• Solar ponds

904 Pad 

West Spray Field 

East Spray Field 

North Spray Field 

East Trenches 

When the 207 Solar Evaporation Ponds were closed, sludge was mixed with concrete to form

blocks. The salt brine and evaporator salts from Building 374 were also mixed with cement to 

produce saltcrete. The 750 and 904 Pad were used to store pondcrete and saltcrete. There were 

immediate issues with container integrity, sloughing, and runoff to the ponds (Norton et al. 

1992).

The potential exists for leachate release as sloughing of waste materials could be 

caused by exposure to precipitation and temperature changes with attendant 

contamination of the area adjacent to the storage pad... 

Independent inspectors from the Nevada Test Site noted that tarps protecting the concrete were

punctured or torn, boxes had suffered significant water damage, and pondcrete boxes were 

bulging at the sides (Norton et al. 1992). By August 1988, water samples taken around the pads

were showing both alpha and beta contamination. 

In the 1980s, treated sanitary

to B-3 Pond and subsequently spray irrigated onto fields. The sewage sludge was considered to 

be low-level radioactive waste. As a result, contamination was released into the surrounding soil

and groundwater (Norton et al. 1992). It would therefore seem appropriate to consider soil 

contamination areas other than the 903 Drum Area Pad. Prior to excluding potential exposure 

from other soil contamination

industrial area prior to remedi

themselves or the resulting ru

spread from these areas. The lack of information on these sites in the environmental TBD makes 

it difficult to determine whether assumptions are bounding. Understanding the extent of soil 

contamination would lead to a better understanding of environmental exposure and impacts of

resuspension on dose to workers. 
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en-3, natural thorium, enriched and depleted uranium,

Pu, Pu and 241Am. The results of the Phase I study identified plutonium 

th respect to off-site exposures.

D did not consider

leases from Pu, noble gas releases from Building 886, Cu, and Np (Rope et al. 1999; 

ChemR one

around

The TB le 2004, pg. 28) indicated that the resuspension of Pu 

deposited in the soil from 903 Drum Storage Area suspension releases was included in the 

ontribution of Am to the total dose appears contradictory to the TBD conclusion 

n page 51 that 241

Am was an insignificant contributor to inhalation dose and was eliminated 

from fu

NIOSH Cs, and 

other ra measurement of 

37

Cs in the buffer zone indicates that there was an operation onsite releasing this radionuclide. 

ed

7

nd 1969, and suspension of plutonium-contaminated soil from the 903 Area during unusually 

high-w s from

RFP pl ailure in

SC&A also believes that resuspension of 239+240Pu and 241Am, as well as other radionuclides,

throughout the site could be an important contributor to ambient dose for both monitored and 

unmonitored dose. 

 

5.9.2 Other Radionuclides 

The TBD relied on the radionuclides of concern identified during the Historical Public 

Exposures Studies on Rocky Flats. The TBD states the following (McDowell-Boyer and 

Little 2004, pg. 8): 

The Phase I study identified several radionuclides as potentially significant 

releases – Hydrog

239+240 241

as the primary material of concern wi

Although plutonium is considered the primary material of concern, it was noted that some

radionuclides present at the site were not considered in the TBD. The TB

238 244 237 re

isk 1992 and 1994a). As previously mentioned, 137

Cs was identified in the Buffer Z

the industrial area.

D (McDowell-Boyer and Litt 239+240

comprehensive evaluation of exposure to 239+240

Pu released from RFP for 1964–1969. A timedependent factor approach was used to address the resuspension of soil contaminated by 

released 239+240Pu and deposited as a result of the continuous and discrete events. The TBD,

however, did not indicate that 241Am resuspension was included in calculating the dose, since 

241Am is present, on average, at up to 0.30 times the activity of 239+240Pu. The potential

significant c 241

o

rther consideration.

should consider internal and external dose impacts from 238Pu, 244Cm, 237Np, 137

dionuclides determined to be present in environmental samples. The

1

This should be investigated. 

5.9.3 Episodic Releases 

SC&A has also alerted NIOSH that between 1953 and 1977, five major accidents were identifi

in Figure 6-1 of the ChemRisk report (ChemRisk 1992). On page 28, the TBD (McDowellBoyer and Little 2004) listed some events, including the releases from glovebox fires in 195

a

ind events in 1968 and 1969. In that figure, the gross alpha radioactivity emission

utonium facilities were shown to range from 500 μCi for the Control Valve F
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1974 to rom

Carmin e in 1965

and one t

address TBD in

the list ions be reexamined further. 

ately consider the impact of episodic 

leases, including the five major accidents at RFP and the hundreds of small fires involving both

uranium

5.9.4 Data Adequacy and Completeness 

The en o te ambient 

airborne do

ttachment 4A describes the use of atmospheric dispersion model results from the 

964.

Du

allo

The ma imum annual inhalation intake values for 0Pu for the period 1953–1964 range from

5.68E- C ange

from 7. E s in

earlier yea Flats,

the Colora

following: 

he e 1950s and 

roof

n

fau

Based o C erived by

atmosph ri econsider the

inhalation

4,000,000 μCi from the 903 Pad Oil Leakage in 1957. In addition, in a memo f

e Plott to Larry Goldren from Rocky Flats on July 11, 1986, two explosions (on

in 1969) were identified as two of five major accidents. These accidents are no

ed in the environmental TBD. They are briefly included in the site description

of incidents. SC&A recommends that these emiss

There has been no consideration in the TBD of releases from incinerator fires, ventilation system

fires, the 374 fires occurring in the plutonium area between 1955 and 1971, or uranium fires 

(U.S. DOE Archives). These fires should also be considered in the context of occupational 

internal dose. The environmental TBD does not adequ

re

 and plutonium. 

vir nmental TBD relied on atmospheric dispersion modeling to calcula

se. It was stated (McDowell-Boyer and Little 2004, pg. 10): 

A

Phase II study to estimate air concentrations and intake between 1953 and 1

ring this period of time results were not readily available and usually did not 

w derivation of 239,240

Pu. 

239+24

x

3 p i/year to 346 pCi/year (including the 1957 fire). The values for 1965–2002 r

68 -1 pCi/year to 1,320 pCi/year. It is reasonable to believe that routine release

rs were higher than those in later years. In an update on health studies at Rocky

do Department of Public Health and Environment (CDPHE 1997) stated the

T largest routine releases from Rocky Flats facilities occurred in th

1960s primarily from the Building 771 exhaust stack and Building 776-777

ve ts. These releases were caused primarily from operational problems and 

lty air filtration systems, which were improved in the late 1960’s.

n DPHE data, it would seem that the annual inhalation intake values d

e c modeling may have underestimated actual releases. The TBD should r

intake values for 1953–1964.

From 1965-2002, the estimate of the average intake of 239+240Pu and 241Am via inhalation was 

based on air monitoring data. NIOSH indicates in the TBD that annual intakes derived for these

years are biased towards higher concentrations (McDowell-Boyer and Little 2004, pg. 11): 

Table 4-2 lists estimated annual intakes of 239,240

Pu and 241Am between 1965 and 

2002. The average values for

239,240

Pu in this table are based on measured 
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ted for samplers in that area and not for samplers in other 

RFP areas. Therefore, the average values are inherently claimant-favorable 

Follow ment 

Team t eficiencies were

identifi am, including the ambient air monitoring. In 

989, DOE indicated the following in its own report: 

atmosphere (page 2-7).

).

concentrations at samplers across the site, but often biased toward the highest 

concentrations, typically near the 903 Area after 1970 because concentrations 

are more often repor

when used as an average for the industrial area. 

ing the FBI raid on June 6, 1989, the Department of Energy formed a Special Assess

o independently review the operations and practices at RFP. Several d

ed in the environmental monitoring progr

1

• Deficiencies in Rocky Flats Plant (RFP) effluent air monitoring program may 

adversely affect determination of radioactive materials released to the 

• There are deficiencies in the ambient air monitoring program for 

radionuclides. As a consequence, the accuracy of measured concentrations of 

plutonium in ambient air is questionable (page 2-11).

• The quality assurance and quality control practices for radiochemistry 

analyses in the Building 123 HS&E Laboratory do not conform to generally 

accepted practices. Consequently, the laboratory cannot adequately verify 

the validity of analytical results (page 7-9

• The quality assurance/quality control practices in the Chemistry Standards 

Laboratory (CSL) may not be adequate to document the validity of the 

radiological reference standards prepared by the laboratory. This laboratory 

prepares reference standards for use by the environmental radiological 

laboratories, for instrument calibrations, as internal tracers, and for the 

preparation of control standards (page 7-13).

• The quality assurance and quality control practices for radiochemistry 

analyses in the 881 General Laboratory may not be adequate to document the

validity of the analytical data (page 7-15).

• Land in the vicinity of and to the east of the 903 Pad is known to be 

contaminated with low levels of plutonium and americium-241, but the 

contamination has not been well characterized (page 7-21).

The report further indicates that ambient air monitoring samples are not analyzed for 241Am and

uranium, which can be released independently or in combination with plutonium. With the 

uncertainty in the environmental monitoring data, and the absence of data at some areas onsite,

the NIOSH statement that intake values derived in Table 4-2 are claimant favorable is 

questionable. Certainly, further investigation into the above issues is needed prior to coming to 

this conclusion. 
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ne

ay not be properly located. The existing radioactive ambient air 

monitoring network was established following limited siting criteria which used a 

urther investigation into the representativeness of these air sampling data to the unmonitored 

height

n average value resulting in non-claimant-favorable assumptions. The CDPHE, in its Summary 

m the 

fire, 37 mCi for the 1969 fire, and 3.1 

i for the releases from the 903 Drum Storage Pad. 

he reliance on the air monitoring system and atmospheric modeling to calculate annual 

cies

has

f

m the RFP during its operational period, which ran from 1952 to 

994. The TBD (McDowell-Boyer and Little 2004) indicated that the Phase I study identified 

al

NIOSH has assumed that the air sampling locations are representative of the average airbor

concentration breathed by a worker. In its assessment of environmental conditions, DOE stated 

the following (DOE 1989): 

Samplers m

Colorado State University wind tunnel study (A-21) as a basis; however, the exact 

locations of these samplers have not been recently documented to be 

representative for the Rocky Flats area. 

F

workforce should be considered. Also, it is important to note that the air monitors are at a

of 6 feet and may not represent air concentrations from soil resuspension adequately. 

Incidental releases determined by the state of Colorado contain a range of release values for the 

1957 and 1969 fires that indicate the TBD does not provide a bounding scenario, but instead uses 

a

of Findings published in August 1999, indicated that the 1957 fire plutonium releases were 

between 2.9 and 39 Ci with a median value of 20 Ci. The releases from the 903 Area were 

estimated to range between 1.8 and 15 Ci with a median value of 3.7 Ci. The releases fro

1969 fire were estimated to be between 10 and 60 mCi, with a median of 20 mCi. (CDPHE 

1999). The TBD assumes a release of 21 Ci for the 1957

C

T

inhalation intakes is questionable. The atmospheric modeling appears to underestimate the 

releases from the site, as compared to recalculated routine releases by the CDPHE. Deficien

in the environmental monitoring program may influence the release values derived from ambient

air monitors. Methods to effectively ascertain inhalation intake values for radionuclides other

than 239+240

Pu should be included in the TBD. Reinvestigation into the annual inhalation intake 

values is warranted.

5.9.5 Timeline Needed for Phases of Operations, and Data Availability and Types 

The RFP TBD for Occupational Environmental Dose (McDowell-Boyer and Little 2004)

attempted to estimate the occupational environmental dose, which is defined as the radiation 

dose received in the course of work duties, outside plant buildings, but on the RFP site. In 

Section 4.2.1, two phases of studies, Phase I and Phase II, provided the basis for the analysis o

releases of radionuclides fro

1

several radionuclides as potentially significant releases — tritium, natural thorium, enriched and 

depleted uranium, 239+240Pu, 241Pu, and 241Am. The results of the Phase I study identified 

plutonium as the primary material of concern with offsite exposures. However, the references

cited on page 19 (ChemRisk 1994a and ChemRisk1994b) have defined Phase I as the operation

period between 1952 and 1989. 
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e

hemRisk 1994b) also provided dose assessment values for the RFP. Phase I is time dependent 

ume

4)

,

ons of the phases of production and studies requires additional 

xplanation. SC&A, therefore, recommends presenting a timeline of these phases in conjunction 

ases

ommends that these relationships be made clearer 

rough the RFP operational and post-operational years.

oyer and Little, pg. 10), 

dicates that the estimated annual intake of Pu between 1953 and 1964 was based on 

53 and 1964,

ir monitoring results were not readily available. The report stated that the calculated intakes 

ic 

eter (AED), which is an upper limit for respirable particles (Rood and Grogan 

999, pg. iv), as listed in Table 4-1. The report went on to state that this likely includes particles 

Pu 

hould explain the terminology as it relates to AMAD and the differences in 

commended particle sizes. Consideration should be given to the plutonium particle size and 

solubility as a result of releases from fires. 

The TBD (McDowell-Boyer and Little 2004) did not, however, indicate that the referenc

(C

from the onset of production at the site in 1952 to 1989. Therefore, SC&A would then ass

that only 3 years of the production phase plus 8 years of post-production phase are included in 

Phase II studies. 

In Section 4.2.2, the occupational environmental dose TBD (McDowell-Boyer and Little 200

identified another two phases—the pre-1993 phase, when production activities were ongoing

and the post-1992 phase, when production activities had ceased. Releases were more likely to

occur as a result of past contamination or decontamination and decommissioning activities. 

The overlap in the definiti

e

with data types and availabilities, as well as data sources used in support of the TDB 

(McDowell-Boyer and Little 2004). The report was not clear if the definitions of the two ph

in Section 4.2.1 are the same as in Section 4.2.2. 

In addition, it seems that there is an overlap in the cutoff date (1992 and 1993) between the 

phases. No explanation was given for the basis of that overlap. In addition, the operational year

in Section 4.2.1 was identified to have run until 1994, while Section 4.2.2 indicated that 

production ceased after 1992. SC&A rec

th

5.9.6 Particle Size 

The Occupational Environmental Dose TBD, Section 4.2.3 (McDowell-B

239+240 in

atmospheric modeling described in Attachment 4A. During the period between 19

a

represented an average from six computational nodes in the industrial area of the site. The 50th

and 95th percentile estimates were associated with particles less than 15 μm aerodynam

equivalent diam

1

that are larger than those in International Commission on Radiological Protection ICRP 66 

(ICRP 1994), which has a default respirable size of 5 μm Activity Median Aerodynamic 

Diameter (AMAD), and therefore, the use of the intakes in Table 4-1 is likely to be claimant 

favorable. On pages 37 and 38, Figures 4A-3 and 4A-4 predicted the annual average 239+240

concentrations in air as a function of time for particles less than 30 μm AED in the perimeter 

area surrounding the site. 

The TBD (McDowell-Boyer and Little 2004) did not explain the relationship between the 15 μm

AED and the 5 μm AMAD. The predictions and models dealt with particles less than 30 μm

AED. The TBD s

re
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Constructed Buildings

d

r

the

, due to the large spatial variation in soil, and thus air 

ontamination at the site. 

heights

l

blems. 

in

C&A believes that a claimant-favorable approach is to use, in a prorated fashion, the maximum 

tration

and

s of annual intakes were applied in 1965–2002,

nd if the 50 percentile or the 95 percentile values in Table 4-1 were used for dose 

ilable data as the basis for filling in missing bioassay, 

ir sampling, and external dosimetry data for claimants with missing data. This strategy 

stribution

s a surrogate for missing data. As we have discussed in the past, we believe the latter approach

ch,

5.9.7 RATCHET Use is Suspect in the Vicinity of Closely

Page 28 of the occupational environmental dose TBD (McDowell-Boyer and Little 2004) 

explains the reasons for choosing the RATCHET model for deriving outdoor onsite exposures.

The model can incorporate spatially varying meteorological and environmental parameters, an

includes modules that perform random sampling of the meteorological parameters, allowing fo

Monte Carlo analysis of uncertainty. SC&A is concerned that the use of RATCHET for the 

onsite dispersion calculations at the RFP site, which is characterized by closely constructed 

buildings, would bias the results. The TBD, on page 30, cautioned about the reliability of

method by indicating that predicted average onsite concentrations might not adequately represent 

actual concentrations of interest

c

Examination of Figure 4A-3 on page 37 highlights SC&A’s concerns that average onsite 

monitoring data for the years from 1972 onward were higher than the 95th percentile values 

estimated using RATCHET. This could be expected because the main goal for RATCHET was 

offsite dispersion modeling. Near-field application of the code is not verified. Building

in close proximity could produce an unexpected air flow affecting the results. A wind tunne

experiment to model the building wakes and air flows could be used to correct these pro

However, in the absence of such an experiment, the use of the peak measured value identified

Table 4-2 would be claimant favorable. 

S

recorded data for the years that there were no data. Applying the maximum intake concen

as given in Table 4-2 in the occupational environmental dose TBD (McDowell-Boyer

Little 2004, pg. 17) to the pre-1965 period would be claimant favorable.

5.9.8 Need to Clarify if Maximum Estimates of Annual Intakes are Applied 

It is not clear if the maximum or average estimate

th th a

reconstruction. 

In reference to work recently completed by the National Institute for Occupational Safety and 

Health (NIOSH) on the revised TBDs for Mallinckrodt and Bethlehem Steel, NIOSH has 

adopted the 95th percentile value from ava

a

represents a revision to previous strategies in which NIOSH often applied the entire di

a

can be characterized as claimant neutral, while the former approach appears to us to be more 

claimant favorable and more in keeping with the letter and intent of 42 CFR Part 82. As su

we believe that the “95th percentile” approach be adopted in this TBD (McDowell-Boyer and 

Little 2004) also.
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ection 3 of the RFP TBD for Occupational Medical Dose, ORAUT-TKBS-0011-3 (Furman and 

f

data

ch on

er

CRP Report 34 (ICRP 1982) and NCRP Report 102 (NCRP 1989). This 

an important issue in that doses for all sites are based upon this research and not any 

e TIB (Kathren and Shockley 2005) reasonably evaluates most parameters involved in 

ent medical exposure of workers during the peak years of 

950 to 1980, when most medical worker exposures occurred. Therefore, in an attempt to fill the 

dditionally, the contractor draws heavily on the NIOSH-OCAS (OCAS 2002a) guidelines to 

ical occupational exposures. Review of 

e subject TBD has resulted in a number of findings and supportive issues that are important for 

he current guidelines, as presented in OCAS 2002a, go a long way in assuring that occupational 

. The

t,

5.10 ISSUE 10: OCCUPATIONAL MEDICAL DOSE 

S

Lopez 2004), makes reference and draws rather heavily on prior documents regarding medical 

exposures at the Savannah River Plant (ORAU 2003), and uses ORAU 2003 and the ORAUOTIB-0006, Dose Reconstruction from Occupationally-Related Diagnostic X-ray Procedures 

(Kathren and Shockley 2005), as a reasonable basis for assumptions regarding its estimation o

worker medical exposures at the RFP. The SC&A review of the TBD recognizes the lack of

and formal protocols through the mid-1980s. One cannot attest to the completeness and 

accuracy of the data, but rather the review focused on the relevancy of assumptions derived from

other site profiles and the TIB. 

Preliminary dose estimations for occupational x-rays are derived principally from the resear

radiation doses in diagnostic x-ray procedures (Lincoln and Gupton 1958), which evaluated 

abdominal and spine x-ray dosimetry at the ORNL. Additional calculation of dosimetry to oth

organs is derived from I

is

measurement of dose at the actual site using unknown protocols. NIOSH and its contractor 

(ORAU) have further assessed medical exposures at other DOE sites, and present their 

conclusions in summary in ORAUT-OTIB-0006, Dose Reconstruction from OccupationallyRelated Diagnostic X-ray Procedures, recently published as Revision 3, dated August 2, 2005 

(Kathren and Shockley 2005). 

Th

assessing the potential medical exposure of DOE site workers. In reviewing the RFP 

occupational medical dose TBD (Furman and Lopez 2004), it is apparent that little, if any, 

written records remain to fully docum

1

gap in assessing medical exposures, the contractor applies the assumption in the TIB (Kathren 

and Shockley 2005) as its basis documentation. 

A

establish its assumptions regarding the magnitude of med

th

NIOSH to consider in order to ensure that the reconstructed doses are claimant favorable. These

findings and issues are discussed below. 

5.10.1 Guidelines Needed on What Constitutes Occupational Medical Exposure 

T

medical exposures are included in determining the overall dose estimations for claimants. 

Unfortunately, interpretation of the guidelines has not necessarily been claimant favorable

occupational medical dose TBD (Furman and Lopez 2004) assumes an interpretation which has 

been also considered and applied at other sites, such as the Savannah River Plant. To this exten

the assumption that medical procedures are limited to only one pre-employment and other 
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imate 

support documentation, ORAU-OTIB-0006, Revision 3, (Kathren and Shockley 2005), it is 

also

and

ot

the

tion of embedded metallic 

adioactive materials associated with the milling and machining accidents that occurred with 

s per year 

re probably limited to a small fraction of high-risk workers. To the contrary, there is ample 

s, as part of the

nnual physical at a minimum. The majority of workers had routine chest x-rays at DOE sites 

being 

exposure to those exams described above, and to assign all other exposures as part of 

orker background.

man and Lopez 2004) does not address the potential use of 

her forms of diagnostic radiography to support medical injury diagnosis. This may involve use 

nal medical TBD (Furman and Lopez 2004) discusses medical x-ray exposure as 

eing chest radiography. In the case of chest radiography, it assumes generally accepted 

ibrated

potential annual chest x-rays, as part of routine physical exams, may substantially underest

worker medical exposure when evaluating the totality of occupational medical exposure. 

In

concluded that other examinations may be included, such as special job exams (e.g., respiratory 

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

medical TBD (Furman and Lopez 2004) does not recognize this change in the TIB, and

incorrectly assumes that special chest radiography for respirator certification, beryllium

asbestos workers, and food handlers are accomplished as part of the annual physical. This is n

documented in past medical protocols and often was performed separately at the request of

Health and Safety Department, up through the mid-1980s. 

Another factor not discussed in the TBD is the possible use of x-ray procedures for special 

screenings, evaluating the result of injury and trauma, and the detec

r

reasonable frequency. 

The TBD (Furman and Lopez 2004) makes the conclusion that one to two chest exam

a

evidence that chest x-rays were provided on a voluntary basis to nearly all worker

a

until the mid-1980s, when Federal guidelines warning against routine screening were first

enforced. 

 

Discussions with NIOSH personnel revealed that it was their decision to limit occupational 

medical

w

SC&A believes that it is not claimant favorable to limit occupational medical exams to one or 

two chest x-rays annually, unless medical records and protocols clearly limit the use of 

radiography to a small fraction of workers, which was not the case up to the mid-1980s. 

5.10.2 Potential for Other Types of X-ray Exposures 

The occupational medical TBD (Fur

ot

of isotopes, sealed sources, etc. The TBD is also deficient in that it does little to catalog the 

number, types of x-ray equipment, frequency of use, etc. Little information exists regarding 

protocols to govern the utilization of x-ray units. 

The occupatio

b

protocols as the norm, with little evidence of rigor or protocols being applied prior to 1985. The 

TBD provides no documentation to support the assumption that units were routinely cal

and maintained. The TBD concedes that no attempt to calibrate or measure output of x-ray 
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e RFP to perform routine and preventative maintenance 

uring the 1950 to 1980 timeframe suggests that routine maintenance of x-ray units is not likely, 

t to evidence 

aintenance, calibrations, etc. The lack of defined protocols and basis for approval of 

unit,

ere not operated at greater than 80–90 kVp. 

ccupational

e

in that workers could essentially request an x-ray. The 

equency of screenings and the numbers and types of workers receiving x-rays varied from site 

ay

 a 

ed

special

e RFP for nearly 20

ears suggests their potential heavy utilization, far more than any other DOE site. The PFG unit 

by conventional

ily,

metropolitan hospital using highly trained technicians and well-maintained x-ray and processing 

equipment occurred prior to 2001. Rather, the protocols to supplant dose calculations are 

derived from historic and very controlled studies performed to support dosimetry presented in 

ICRP 34 (1982) and NCRP 102 (1989). 

 

The less than average performance at th

d

unless performed by an outside contractor. Unfortunately, no records exis

m

radiography procedures suggests that the use of radiography was not closely controlled. The

occupational medical TBD (Furman and Lopez 2004) does not discuss the use of portable 

radiography to perform screenings and the potential for exposure of medical personnel or other

workers without dosimetry devices being utilized. This is potentially an issue for the PFG

which was often van-mounted at other sites. Additionally, the TBD fails to document that 

available x-ray units w

The TBD (Furman and Lopez 2004) does little to document the variety of medical o

exposures, the type of equipment, and the type of maintenance performed. In light of this, ther

is a need to reconsider the approach for reconstructing medical radiation exposures. 

5.10.3 Frequency and Types of X-ray Exposure is Derived from Other Sites 

The occupational medical TBD (Furman and Lopez 2004) relies on assumptions of x-ray 

frequency derived from other DOE sites. The assumption of one to two chest radiographs per 

year is not reasonably conservative,

fr

to site. 

The occupational medical TBD (Furman and Lopez 2004) provides no documentation or 

references to support the assumption that only a limited group of workers received annual x-r

exams. To the contrary, up until about 1985, most DOE sites performed chest x-rays on

voluntary basis. DOE medical program reviews documented during the early 1990s showed

many sites still used chest radiography as a general screening exam. Most workers accept

chest x-rays, even though the job did not require it. Also, the assumption that workers in

exposure categories, such as beryllium workers, were given chest x-rays only as part of their 

annual physical is not documented. 

 

The occupational medical TBD (Furman and Lopez) does document that photofluorography 

(PFG) units were available from 1953–1968. The presence of PFG units at th

y

also provides a dose to the worker greater by a factor of 5–6 than that delivered

radiography. 

Also, the ORAU-OTIB-0006 (Kathren and Shockley 2005), upon which the TBD relies heav

uses retake rates that average about 3%. The study referenced was based upon a large 
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icians (Federal regulations did not require technician 

ertification), showed that retakes sometimes ran up to 30% for abdominal exams and often over 

etermination of Machine and Technician Uncertainties 

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

factor of 1.3

claimant favorable. 

. That

udy was limited to abdominal and spine x-ray procedures, using vastly different x-ray 

m derived

from ICPP 34 (1982) and NCRP 102 (1989), and are not reflective of equipment and protocols 

sed at the RFP during 1950–1975.

ngs, etc.) all contribute to overall uncertainty 

lated to dose calculations. In an attempt to corroborate, Dr. John Mauro, in a private 

ith the New Jersey Department of Radiation Control, found the following 

formation regarding overall uncertainty in x-ray machine output (Mauro 2005). Using 

standard deviation often exceeded the mean ESE in that study. Given that the New 

rsey assessment involved improved technically enhanced x-ray equipment, and followed an 

esta

reas

The oc an and Lopez 2004) does document that 

pho

units were not considered when organ dose calculations were documented in ICRP 34 (1992) and 

NC TBD

oc exam, 

units. A comparison review of Federal facilities, such as by the Department of Defense (DOD) 

during the 1970s using lesser-trained techn

c

15% for chest radiography. Although the PFG, by design, is less likely to require retakes, it is 

inherently much more dose-intensive. 

5.10.4 The D

The occupational medical TBD (Furman and Lopez 2004) provides little documentation to 

su

derived from ORAU-OTIB-0006, Revision 3 (Kathren and Shockley 2005). Additionally, the 

TBD does not address the potential impact of associated uncertainties, nor does it provide 

documentation to warrant its assumption that multiplication of estimated doses by a

is

The occupational medical TBD (Furman and Lopez 2004) is deficient in that little 

documentation exists to validate x-ray protocols, equipment maintenance and upkeep records 

prior to 2001. The TBD uses information derived from the Savannah River Site review (ORAU 

2003) and from Kathren and Shockley (2005) to estimate dose impacts. Further dose estimations

are seemingly derived from a study performed by Lincoln and Gupton (1958) at ORNL

st

equip ent, which could substantially impact dose calculations. Actual organ doses are

u

The TBD estimates an “uncertainty factor” of 30% to account for variability in machine output, 

due to fluctuations in voltage, amperage, and timer accuracy; however, other factors that 

contribute to uncertainty are overlooked. For example, contributions due to technician errors 

(i.e., wrong techniques, improper machine setti

re

communication w

in

established protocols, New Jersey reported that the entry skin exposure (ESE) for chest 

examinations from nearly 500 units over a 3-year period varied by up to 2 orders of magnitude, 

and that the

Je

blished protocol, we believe that application of the 1.3 uncertainty factor in the TBD is not 

onable and does not assure the dose determination is claimant favorable. 

cupational medical TBD (Furm

tofluorography (PFG) units were available to be used at that site for nearly 20 years. PFG

RP 102 (1989), which forms the basis for organ dose calculations presented in the

d ument. Kathren in the ORAU-OTIB-0006 does set an exposure of 3 rem per PFG
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prior

lso, the occupational medical TBD (Furman and Lopez 2004) does not consider the importance 

e on retake rates and diagnostic quality of 

e exam. Improper processor maintenance will also appreciably affect the overall dose 

ts

he occupational medical TBD (Furman and Lopez 2004) does not show that the RFP applied 

dose m sess or

conside ollowup. Th bly show or

match ent the 

numbe ds. 

dose reconstruction. These missed doses may have occurred as a result of incomplete 

.11.1 General

g their

ose reconstruction methodologies and assumptions to address a number of issues, including the 

followi

(1) RP data

(2) use in

(3) 

(4)

which is considered reasonable. Therefore, NIOSH should assure that all chest exposures

to 1964 are based upon a 3 rem estimate. 

A

of processor maintenance, and the impact it may hav

th

uncertainty.

5.10.5 Use of Screens, Grids, and Impact of Off-Site Medical X-rays are Not Considered 

Finally, the occupational medical TBD (Furman and Lopez 2004) does not consider dose impac

due to less than optimal use of technology, such as using screens, grids, or bucky systems. The 

TBD does not also consider off-site medical exposure, as part of worker exposure. 

T

inimization principles to reduce medical exposures. The document also does not as

r the likely exposure to workers who are referred to off-site medical facilities for f

e TBD states that review of selected medical records and files did not reasona

expected x-ray exam frequency, and type of exam. Little evidence exists to docum

r of x-ray exams provided to the average worker or for special exposure nee

5.11 ISSUE 11: COMPLETENESS AND ADEQUACY OF DATA FOR DOSE 

RECONSTRUCTION 

A considerable amount of evidence exists that there is a potential for missed occupational 

external, internal, and environmental dose due to the completeness and quality of data being used

in

monitoring records and inadequate monitoring techniques. There may also be a data integrity 

issue associated with the external dose record.

5

Based on the conference calls with the TBD authors, NIOSH is in the process of updatin

d

ng:

Developing a technical information bulletin to provide direction on the use of ND

Acquiring the Rocky Flats Job-Exposure Matrix put together by Dr. Ruttenber for

determining co-worker doses 

Developing a technical information bulletin dealing with dose calculations from highfired oxides 

Considering the inclusion of additional incident reports in the site description 
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Althou

dosime te

future r

.11.2 Completeness of External Exposure Data 

ith external radiation monitoring at the RFP in the early years (i.e., 

951 through the early 1960s) was because of the general belief at that time that there would not 

ome

TSAs found that procedures for calibrating radiation monitoring instruments 

were weak and poorly documented. The accuracy of instruments is important to 

Withou ether the

ealth physics program was effectively identifying those individuals with the potential to 

orkers

-ray

eters were used to monitor both whole-body and wrist exposures. 

nly certain groups of workers were issued dosimeters until 1964. By 1964, dosimeters were 

rs to be exposed to radiation

without being monitored in accordance with the Radiological Control Manual 

ount of the workday 

in the area. RFP personnel noted that a guard station was adjacent to the area, 

and that the guards were not required to wear dosimeters on a routine basis. The 

gh these revisions to the TBD are in progress, the following review of the external 

try program is included in our review to serve as a baseline against which to evalua

evisions of the TBD.

5

Many of the inadequacies w

1

be a great deal of external radiation hazards associated with the operations at the RFP (Putzier 

1982). Based on this belief, external monitoring of workers was limited to workers who were 

judged to have the potential to experience 10% of the radiation protection guide (RPG) at that

time. Furthermore, implementation of this external radiation monitoring strategy was hampered

by the inaccuracy of radiation monitoring instruments and the lack of radiation surveys in s

areas. The GAO report elaborated on calibration issues at RFP (GAO 1988), as follows: 

The

ensure that workers and the public are not exposed to unnecessary levels of 

radiation. Problems included using different calibration techniques in different 

buildings and not testing instruments with appropriate radiation sources. 

t appropriately calibrated instrumentation, uncertainty is introduced as to wh

h

experience greater than 10% of the RPG. For example, the plutonium metal (foundry) w

in Building 771 were not monitored for whole body, penetrating gamma radiation and x

doses until February 1957. Instead, they were issued only a wrist dosimeter (Falk et al. 2005). 

From 1953 to 1970, film dosim

O

incorporated into the security badges of all employees at RFP (Inkret). 

The DNFSB has expressed a concern regarding Building 771 that personnel outside the 

immediate area may have been exposed to radiation without the benefit of monitoring. 

The staff noted that a potential existed for worke

(RCM) and DOE Order 5480.11. In discussions with Building 771 personnel, it 

was noted that the Thermoluminescent Dosimeter (TLD) badge storage rack was 

being evaluated to determine the amount of radiation the dosimeters were 

exposed to while hanging on the rack. This evaluation was being accomplished 

as a result of RFP personnel noting that two TLDs that hung on the rack for six 

months had received approximately 300 mrem. The DNFSB staff questioned 

whether any unmonitored workers had spent a significant am
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ew, and was

ilar to that at the 

TLD storage board, exposure of guards to ionizing radiation may exceed the 

s a result of the lack of external monitoring in some worker populations, there is a potential for 

where the

amma component of external dose equivalent exceeded certain criteria, individuals were also 

e

he

assigning secondary dosimeters, which 

imetry was the responsibility of the field 

ho

to the

reconstructors used dosimeter measurements. The 

BD d s. ORAUTTKBS600 mr at the

upper 9 For

those in assigned

(pg. 43

radiation level in the guard area was not known at the time of the revi

to be determined. If the radiation level in the guard's post is sim

100 mrem per year limit for those who are not monitored. (DNFSB 1993)

A

missed beta/gamma dose. 

Lack of neutron monitoring also introduced gaps in the monitoring data. Prior to 1957, 10–18 

individuals involved in final assembly operations received neutron dosimeters. In 1957, a Los 

Alamos National Laboratory scientist discovered the substantial neutron exposure associated 

with the fluorination process. Starting in the period between 1959 and 1960, everyone working 

in Building 771 was monitored for exposure to neutrons. In other areas at RFP,

g

monitored for exposure to neutrons (Inkret). With respect to neutron monitoring, GAO stated th

following: 

When there was initial concern in the mid 1950’s that personnel at Rocky Flats 

were receiving neutron exposure above 10% of the allowable standard (5 rem per 

year at the time), neutron track plates and/or later NTA film were assigned to 

some personnel depending on their work locations. These early efforts to identify 

potential neutron exposure problems were not well explored (Baker 2002).

Some of the highest exposed individuals were not monitored during their early work in 

fluorination and metallurgy processes. The Neutron Dose Reconstruction Protocol has 

reassessed neutron doses (Falk et al. 2005). NIOSH/ORAU has received these data and is in t

process of developing a technical information bulletin on how to incorporate these data into the 

dose reconstruction process. 

 

Even as late as the 1990s, RFP had not established a procedure for identifying the workers that

required extremity monitoring. The responsibility for

were not in common use until later, and extremity dos

radiological control organization. There were no consistent criteria for determining w

received extremity dosimetry (DNFSB 1993). Also, the capability for monitoring neutrons

extremities was not available until 1972 (Baker 2002). 

Where dosimetry data is available, the dose

T ivides unmonitored workers into production and non-production worker

0011-6 (Langsted 2004) instructs the dose reconstructor to assign a missed dose of

em per year (5% of the annual Radiation Protection Guideline), or 1.2 rem per year

5% confidence level for unmonitored individuals in the production areas (pg. 42).

dividuals who worked outside the production areas, an environmental dose was

). With respect to extremity doses, the TBD states (pg. 43): 
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odology for assigning missed extremity dose. This is especially true when there 

as no method for neutron extremity monitoring. 

In orde k

location ls

withou badges)

must be carefully evaluated to ensure there was not a potential for occupational exposure (e.g., 

xposure from duct work). For those individuals not entering any radiological areas and having 

data, the approach of assigning environmental dose is appropriate assuming these 

oses take into account all potential modes of exposure and appropriate data. 

NIOSH s and 

reasses ch as

uranium achment 4). 

The ND orkers

include

otenti osed to neutrons, or production and support personnel who entered the plutonium

ns

dicates that neutron dose was not applied in some

ases w the

TBD a

maxim ate the

comple ntially

xpose ction should be provided on when to assign missed neutron dose. 

he inc

inconsi

rogram ting

Additional information on these dosimeters will be required for dose 

reconstruction for shallow dose to the extremity, if necessary. 

There is no meth

w

r to assess the methodologies used above, it is necessary to know the job title, wor

, and tasks an individual was involved in during a particular year. Those individua

t dosimeter data (i.e., during years when dosimetry was not paired with security

e

no monitoring

d

/ORAU indicated that the scope of the NDRP was to reevaluate the available film

s the dose for those individuals only. The scope did not include other workers, su

 workers and workers who were not monitored with neutron film (see Att

RP data will allow dose reconstructors to assign missed neutron doses for those w

d in the study; however, no methodology has been developed for uranium workers

p ally exp

production area but were not provided neutron monitoring (e.g., company auditors). Missed 

neutron dose has not been assigned in all non-compensable cases where maximizing assumptio

apply. Specifically, a review of a few non-compensable cases on the NIOSH Occupational

Claims Tracking System (NOCTS) database in

c here claimants worked in plutonium areas. This indicates inadequate guidance in

nd other supporting documents for application of missed neutron dose, at least in 

um dose cases that are estimated to be non-compensable. The TBD should investig

teness of the NDRP for assignment of missed neutron dose with respect to all pote

e d workers. Clear dire

5.11.3 Completeness of Internal Monitoring Data 

T ompleteness of data was not limited to external monitoring. There was some

stency in the internal monitoring program as well. In a review of the RFP bioassay

p , the DNFSB identified a potential for missed internal dose as a result of not collec

bioassay data, as follows (DNFSB 1994): 

Additionally, it was not apparent to the DNFSB staff that the program 

encompasses all applicable personnel. For example, EG&G personnel described 

that the program's data base is derived from the list of personnel trained on site, 

and would not track visiting workers. In this case, EG&G personnel explained 

that it would be a sponsors responsibility to see that the bioassay had been 

accomplished. 
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d not make clear how contractors are included in the bioassay 

program. For example, when asked about other subcontractors, RFP personnel 

other means. (DNFSB 1993)

Further

ed that the program

has not been fully effective because the procedure has only recently been

In the c

dose. To calculate internal dose for workers not involved in an incident, the internal TBD 

provides the following guidance (Falk 2004): 

The sug the worker in a building

at an arbitrary fraction of the inplant guide (or official limit), whichever is more 

To calc

followi

made.

In t c

radiolo such as

job title, work location, skill or task, time spent in an area, and radionuclides of concern are not 

read r

assigni omplicate the matter. In 

Even when a routine internal monitoring program was in place, there was no mechanism to 

ensure that workers and subcontractors were submitting timely bioassay samples. The DNFSB

noted the RFP had a significant number of bioassay delinquencies that may have led to missed 

internal dose. 

The discussions di

indicated that contractors may come and go and not be bioassayed. A deliberate 

delinquency tracking system does not exist, but delinquencies are identified by 

more, 

For example, EG&G personnel identified that the bioassay delinquency rate is 10 

to 12 percent, and that some delinquencies durations are at a point where 

analysis sensitivity may be impacted. EG&G personnel not

promulgated. The DNFSB staff believes that these delinquencies may be 

undermining the effectiveness of the program, and dose may be missed. (DNFSB 

1994)

ase of an unmonitored worker, there are several techniques used by the TBD to assign 

gested approach is to estimate the time spent by

involved in a radionuclide of interest and credit the worker with a chronic intake 

claimant favorable. 

ulate internal dose for workers involved in an incident, the internal TBD provides the

ng instructions.

Claims files may include event-specific data that should be used to reconstruct 

internal dose. When such data is not available default assumptions may be

he ase of an unmonitored worker not involved in an incident but having access to 

gical areas, it is not clear whether the method outlined is bounding. Information,

ily available and would have to be evaluated for the particular individual. The process fo

ng dose to an individual in multiple job locations would further c

the case of an incidental exposure, not collecting samples in a timely manner may prevent 

dosimetry from collecting needed follow-up samples, or render bioassay techniques ineffective

(e.g., insoluble plutonium). 
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ors,

ant Personnel Dosimetry files has led to some observations 

elated to completeness of records provided to NIOSH/ORAU. 

• Although annual summary information is available, the quarterly dosimeter results are 

coupled with the security badge from 1964 through early-1990s 

(Langsted 2004, Vol. 6, pg. 18). Despite this, blank records were observed in at least two 

• everyone was badged over this period of time, there should not be blank records – that 

itored

se was 

ual

itly explained

sed:

) for the

blank records would also appear to be warranted, especially in view of the concerns 

969 when one of the major fires occurred at the site. 

• amples, yet there was no

dication of an incident or personnel contamination in the file. 

With si gle or ce

personnel, man

cords hould

In summary, there is partial or complete absence of records for many former workers, visit

and subcontractors. As a result, this requires careful consideration of an individual’s job title, 

work location, radionuclides of concern, and acute or chronic exposure potential. These people 

include those likely exposed to radiation and those who may have no exposure potential. 

5.11.4 Dosimetry Files 

A review of several Rocky Flats Pl

r

absent from each of eight files reviewed for 1970. 

• The dosimeter was

cases for the post-1964 period (see also discussion below).

If

is records containing no data whatsoever, not even a zero—in the file of a mon

worker. It appears to have been the general practice to record a zero when the do

below the detection limit, for instance. SC&A has found that, in the 1950s, individ

dose records entitled Health Physics External Exposure Activity Run explic

zeros and blanks as follows: 

A zero indicates a “0” dose received. 

A blank indicates that no film badge was worn. 

If this interpretation is valid for the post-1964 period, the following issues may be rai

• Blank records for monitored workers appear to be problematic and in need of explicit 

guidance for interpretation and dose estimation. An inquiry into the reason(s

regarding data integrity discussed below. 

• Doses for subcontractors may be incomplete or absent. Interviews indicate that these 

individuals were monitored; however, no monitoring records were located. 

• Gaps exist in individual records, and there is no explanation for the absence of data. This 

included gaps for the year 1

There are records where individuals had positive bioassay s

in

n multiple year gaps occurring in records, especially for support and maintenan

y personnel dosimetry files appear to be incomplete. The absence of these 

re s be investigated.
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us

ble to locate any investigation reports as a 

sult of contaminated badges in Rocky Flats dosimetry files. 

n

that out in our most recent report. And DOE is taking a look at that as to ways to 

a

ng this dose notation were commonly in jobs that were associated 

ith higher radiation exposure (e.g., chemical operators, NDA support) (see Attachment 2). A 

R l m Roger Falk of the Health Physics Department explaining the 

nnual employees radiation exposure reports, which was apparently an official routine cover 

as made (Falk 1987): 

2) the exposure was so small it cannot be measured by our current methods 

Dose reconstruction guidance concerning the handling of contaminated, lost, damaged, or 

unreturned badges is not addressed in the site profile. When dosimeters became contaminated,

they were either decontaminated or replaced with new dosimeters. According to health physics, 

when a dosimeter was too contaminated to be processed, a dose was estimated based on previo

dose history or co-worker data. SC&A has not been a

re

 

5.11.5 Data Integrity Concerns 

5.11.5.1 Zero Recorded Doses 

When dosimeters were lost or not returned, radiological control was to perform an investigatio

and assign an estimated dose (see Attachment 2). In testimony to the Congress, the following 

was stated (HCEC 1994): 

Mr. Schaefer: If a dosimeter was not returned, should an estimate of the exposure 

to radiation have been made? What’s the result of not making that? 

Mr. Wells: Well, we think perhaps an estimate that would have been included 

would have certainly been better than to report zero exposure. And we pointed 

correct their existing procedures, and that would include providing better 

estimates of what might have taken place. 

The assignment of zeros or nothing in lieu of an actual estimated exposure for lost, 

contaminated, damaged or overexposed film would result in the dose reconstructor assigning

missed dose which may or may not reflect the actual exposure received. In cases where 

secondary dosimeters were used (not common until later years), this value could provide 

information on the actual dose. Also, radiological survey data, timekeeping data, and co-worker

data can also provide a basis for estimate or validation of the missed dose method. 

Workers indicate that the dose reported to them included entries stating “No Current Data 

Available.” Workers receivi

w

ockwel memorandum fro

a

memo accompanying workers’ dosimetry used for an undetermined period of time, the following

statement w

A 0 (zero) is recorded for any of the following conditions: 

1) no exposure received 

3) no data is available to make this assessment.
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the quote above validates the GAO testimony that zeros 

were entered when no data were available, including the specific instance when the data were 

lacking badge.

he practice of recording zeros when badges were not turned in raises questions about the 

tegrity of the external dose record 

will take considerable analysis of dose records, job types, and area radiation data, coupled with 

deta d

zeros w te as 1994, the entire period of

Rocky Flats operation, as well as at least part of the post-production environmental remediation 

per ,

maintenance personnel, and other job types with potential for exposure to radiation who were 

onitored and have zeros in their records may be open to question until this question is resolved. 

the case of a zero dose, the dose reconstructors are to estimate the number of zeros based on 

logy for

.

his may not always be a claimant favorable. Dow Chemical (1968) indicated in a report 

disc s

plant’s 1,200 plutonium production workers received penetrating radiation doses exceeding the 

AEC’s 5 rem standard in 1966. Dow traced most of the problems to production changes that 

ena d er

exposures in excess of 5 rem. Rockwell International (1987) confirmed these numbers in a 

report on radiation exposures to Rockwell employees. The dose reconstructor should take this 

pos i projects, such as 241Am

purification and ZPPR fuel. 

Blank Records 

he dosimeter was coupled with the security badge from 1964 through early-1990s, according to 

o other RFP production operations. In 1964, the

security badge was incorporated in the dosimetry badge, which ensured that each 

If every

specially since it was the general practice to record a zero when the dose was below the 

An inquiry was made of the author during a conference call with NIOSH in regards to the 

meaning of “no data is available to make this assessment.” Clarification on this issue was not 

provided (see Attachment 4). Item 3 in

because the worker did not turn in the

T

integrity of the external dose record. An analysis of the frequency of this practice and the 

period(s) over which it occurred is essential. This is especially important since workers report 

that exceeding the quarterly limit would result in a downgrading of work assignment or even a 

layoff. This appears to be similar to a problem that existed NTS for some time (Hacker 1994). 

Under such conditions, it is possible that doses were above the quarterly limit. How frequent 

such a practice might have been, and its implications for the in

ile worker interviews (including retired worker interviews). Since the practice of recording 

hen badges were not turned in was reported by GAO as la

iod should be examined. The records of all or some production workers, guards, 

m

In

the dose level and the exposure limits for the year in question. The missed dose methodo

recorded zero doses assumes that individuals did not exceed the Radiation Protection Guideline

T

us ing radiation problems with plutonium fabrication operations that eighty-seven of the 

ble the plant to make more nuclear bombs. In 1967, the plant reported eighty-eight work

sib lity into consideration, especially for those working in special

5.11.5.2 

T

the TBD: 

The use of dosimetry expanded t

individual wore a dosimetry badge (Langsted 2004, Vol. 6, pg. 18).

one was badged after 1964, there should be no blank records in the files of workers, 

e
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A zero indicates a “0” dose received. 

It is un eriod, when all

personn Hence, the existence of

blank r s for monitored workers appears to be problematic and in need of explicit guidance for 

interpre ld also appear

to be w integrity in the

post-19

• ntegrated ID-dosimeter badge after 1964 and, if not, what 

e criteria for issuing only an ID badge might have been 

• de 

.11.6 Additional Issues Pertinent to External and Internal Dose Reconstruction 

There a ring and

nvironmental data that may influence dose reconstruction. The issues include the following: 

• ds of

• Inadequate bioassay techniques existed for the type of radioactive material present in 

• Inconsistent directions were provided by health physics staff 

• tal monitoring techniques for airborne releases were questionable 

5.11.6

One of the issues that raise questions about the adequacy of the external dosimetry records at 

Rocky th

241Am,

alent, it became obvious to me that the simple Oak 

Ridge badge was inadequate (Putzier 1982). 

detection limit. SC&A has found that blank records may indicate no monitoring at all. In the 

1950s, individual dose records entitled Health Physics External Exposure Activity Run explicitly 

explained zeros and blanks as follows: 

A blank indicates that no film badge was worn. 

clear whether such an explanation would have been retained in the post-1964 p

el entering the site were provided with an integrated ID-dosimeter.

ecord

tation and dose estimation. An inquiry into the reason(s) for the blank records wou

arranted, both as an issue that is relevant for dose reconstruction and for data

64 period. This inquiry should include the following:

Whether all personnel wore an i

th

The criteria for and circumstances in which no entry, not even a zero entry, was ma

into dose record 

5

re a number of issues associated with the quality of the radiation monito

e

Dosimetry algorithms, calibrations, and practices were problematic for some perio

operations 

some facilities 

Environmen

.1 External Dose Reconstruction Issues 

Flats has to deal with external exposures to the low energy photons associated wi

as evidenced by the following: 

As I became more aware of radiation properties of plutonium, and also as 241 

americium became more prev
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Even in metry 

calibrat merce 

(HCEC at Rocky Flats

that cou AO stated: 

at is very good. I’m glad we did that. You’re testimony on page 

2 describes problems with monitoring and sampling of air to detect radiation 

xisted 

to calibrate equipment to even tell if the correct dose had been received, were not 

Mr. Schaefer: I understand that there were calibration problems. Does that mean 

as

).

later years, concerns were expressed regarding the adequacy of the external dosi

ion program at the RFP. In testimony to the House Committee on Energy and Com

1994), the GAO was asked whether there were problems with monitoring

ld have led to undetected radiation releases. The witness Jim Wells of the G

Mr. Schaefer: Th

releases at Rocky Flat in 1988. What effect would such problems have had? 

Could the problem have related to undetected radiation releases? 

Mr. Wells: I think the answer would be, it’s possible, because they were unknown 

in terms of what actually was out there in the air and how much was being 

ingested by the workers. The actual dosimetry devices, the program that e

fully functional and operating. 

And, as the chairman pointed out in his opening statement, they were doing all 

kinds of mathematical adjustments for the exposure doses, some of which were so 

artificial it ended up with negative exposure, so the potential is there and they 

could have but no one knows. 

that workers could have been exposed to radiation and not realized it?

Mr. Wells: What that meant was the device that they had pinned on them, that 

provided a reading at the end of the day or whenever it was measured, there w

no assurance that it was reading high, or low, or reading at all. That’s what a

lack of calibration program means. 

In a safety audit conducted in 1986, on the external dosimetry program, there were several

findings related to quality control procedures and calibrations (Rockwell International 1986

There is no formal verification of accuracy of dosimeter results for field 

conditions. 

New instruments and dosimeters are not acceptance tested using beta, photon and 

neutron radiation sources according to their intended field use. 

Instrument and dosimeter response is not characterized for radiation types and 

energies present in the work environment. 

In addition to calibration issues, there were issues associated with dosimeter algorithms. The 

DNFSB indicated in a 1993 review: 
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Laboratory Accreditation Program (DOELAP) in 1990, and that the accreditation 

n

case the Dosimetry Department had to know the facility in which the 

dividual worked (Langsted 2004, pg. 19). This approach to calibration could be problematic 

f unauthorized practices by RFP personnel. Individuals apparently 

verexposed their badges on purpose (e.g., putting the badge in the glove or on a waste 

Attachment 2).

.

ould

s 

n

eter 

RFP personnel noted that the Whole Body Dosimeter Program passed the DOE 

certificate was received in October 1991. RFP noted that the program's 

algorithm has been changed since the last accreditation, and that a wrong 

correction factor may have been applied in the case of the K17 dosimeter chip. 

(DNFSB 1993)

The external TBD indicated that film dosimeters required a work-place specific calibratio

factor. Since this was the

in

for workers whose job took them into various areas of the plant (e.g., security guards, auditors,

engineers, maintenance, etc.) Some discussion on the impact of using the incorrect calibration 

factor should be included and adjustments made to the uncertainty, as necessary. 

There is some evidence o

o

container) to get out of working in a particular area. As a result of this practice, individuals who

actually received high doses were accused of overexposing their badges and were in some cases

disciplined. In order to protect themselves from disciplinary action and/or ensure they were 

eligible for overtime, individuals would leave dosimeters in their lockers or put them in their 

back pockets (see

In a 1988 review by the GAO, the apparent complacent attitude towards safety was addressed

Worker attention to radiological protection needs to be increased. According to 

the TSAs, workers have been observed handling contaminated items without 

surgical gloves, not surveying themselves for contamination, and improperly 

wearing dosimeters (badges used to measure exposure to radiation.) (GAO 1988) 

Improperly worn dosimeters will affect the dosimeter response. Given that not wearing a 

dosimeter or wearing it incorrectly (e.g., in back pocket) can influence the response of the 

dosimeter, the relative difference between default assumptions and actual field conditions sh

be considered and appropriate correction factors determined. Also, the site profile assumes an 

exposure geometry based on the fraction of time an individual is working hands-on with 

material. NIOSH/ORAU need to verify these values with workers and expand the work 

categories to include all types of workers. During site expert interviews, many of the site expert

disagreed with the amount of hands-on time NIOSH/ORAU had assigned to their job title. I

other situations it was not clear which category a worker fit into. 

In a 1992 draft of the RFP external dosimetry procedure, workers using lead aprons were 

directed to wear their dosimeters outside and in contact with the apron (RFP 1992). In the 

Rockwell Health and Safety Manual (RFP 1988), workers were directed to wear their dosim

under their aprons. The details of which practice was observed could influence the 

determination of photon dose to exposed portions of the body. The TBD should make a 

claimant-favorable assumption with regard to dosimeter response and wearing lead aprons. 
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were

. It was not until the 1970s that production 

orkers were put on a routine lung count. There were some individuals onsite who did not 

uper S

and

not

plutonium present at the facility, urine, fecal, and lung counts in 

ombination were needed to effectively identify potential intakes. The greatest uncertainty in 

e

assay results were recorded as background or less than the 

etection level.

tigated the potential importance of this issue by back-calculating the doses to the 

rgans of the respiratory tract that might be associated with levels of plutonium in urine that are 

used to model the retention of Type S and Type Super S plutonium in the body, the 

oses to the respiratory tract could differ by over an order of magnitude. However, no matter 

what a ither Type S

or Typ n urine,

the dos hose of the

respira

een Type Super S, will result in a significant overestimate of the doses to those organs. 

for

e

t because, 

epending on the assumptions employed, the doses to those organs can differ by more than an 

5.11.6.2 Internal Dose Reconstruction Issues Related to the Inhalation of Plutonium 

High-fired plutonium oxide (often referred to as Type Super S) existed in the workplace as a 

result of both production activities and fires. The early bioassay program utilized urinalysis to 

detect intakes of plutonium. In 1964, a lung counter was put into operation. Lung counts

initially used in incident response starting in 1964

w

receive a lung count until the 1980s. Fecal sampling was done only if an individual was 

involved in an incident. 

Urinary excretion does not provide an effective method for detecting Type S or Type S

plutonium using the methodologies employed in the early years of RFP operations. At that time, 

the lower limits of detection of plutonium in urine were poor, and large quantities of Type S

Type Super S plutonium could have been inhaled (both chronically and acutely) and not have 

been detected in urine. 

If there is sufficient 241

Am in-growth in the product inhaled, it is possible that the 241Am can be 

detected during a lung count and used to derive the amount of plutonium inhaled. However, if 

the amount of 241Am inhaled, relative to 239

Pu, is not known, or if the inhaled plutonium does

contain detectable levels of 241Am, this method of estimating the lung burden of plutonium could 

be misleading. 

With the different forms of

c

missed dose is likely for workers who only participated in the urine bioassay program. This was

the case for all workers prior to 1964. There are situations where individuals could hav

received uptakes, but their bio

d

SC&A inves

o

at the lower limits of detection for plutonium. We have determined that, depending on the 

assumptions

d

ssumptions are used to derive the doses to the respiratory tract from e

e Super S plutonium, even at the minimum detectable levels of plutonium i

es to the respiratory tract are extremely high. For organs other than t

tory tract, the assumption that the inhaled plutonium is Type S, when in fact it might have 

b

As discussed earlier, SC&A has concerns regarding the methods used to derive the MDA for 

plutonium in urine and the assumptions used by dose reconstructors for deriving doses

workers when the urine bioassay records indicate zero or background. These concerns ar

important for reconstructing the doses to organs other than the respiratory trac

d
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robability of causation. This issue is not as important when deriving the doses to the respiratory 

tract ba regarding

the MD lt in a high

probab

thoroug

Many o lutonium were resolved with the 

stitution of lung counting as a means for estimating the lung burden of plutonium in workers 

it is first used to manufacture 

eapons, inevitably contains a certain amount of Pu, with the amount dependent on the 

irradiat 14.4 years.

If the w

downstream of that location would have very different chest or body counts compared to workers 

pstream for the same level of Pu intake. Further, even after some time, the levels of 241Am will 

be low ium that 

has bee than in

plutoni s not

only a uires a

history after first

brication into a weapon.

NIOSH e from

241Am

reconst

Interna an 

Pluton

With re 4, pg. 7) states the

ollowing:

ntakes of plutonium, 

americium, enriched uranium, depleted uranium, and tritium, as well as other 

Althou ied as

produc uptakes 

of thes chniques for these

dionuclides and in some cases no discussion. The TBD does not discuss how potential 

order of magnitude and in a dose range that could be significant in terms of the associated

p

sed on urine analysis data because, no matter what assumptions are employed

A, the doses to the respiratory tract are extremely large and are likely to resu

ility of causation. We believe that these are important issues that need to be more 

hly explored and discussed in the TBD. 

f the issues associated with urine bioassays for p

in

based on the amount of 241

Am detected by the lung count. However, there are some issues in 

regard to the adequacy of the record in relation to the use of 241

Am measurements for 

determining plutonium intake. Weapons grade plutonium, when

241

w

ion level of the fuel. Over time, the 241Pu decays into 241Am with a half-life of

arhead is then dismantled and the plutonium purified to remove 241Am, the workers 

u

compared to fresh plutonium that has aged, because the amount of 241Pu in pluton

n subject to 241

Am removal after the return of a weapon would be much lower

um straight from the reprocessing plant. Hence, estimating worker Pu intake require

measurement of 241

Am and knowledge of the age of the plutonium, it also req

of how many times the plutonium has been processed for 241Am removal

fa

should ensure that the appropriate records are available for determining Pu intak

whole-body or lung-counting data for the individual worker for whom the dose 

ruction is being done.

l Dose Reconstruction Issues Related to Exposures to Radionuclides Other Th

ium

spect to radionuclides of concern, ORAUT-TKBS-0011-5 (Falk 200

f

Workers at Rocky Flats had the potential to receive i

radionuclides. ...Site-specific internal dosimetry information for other 

radionuclides, such as thorium, curium and neptunium, is rare or not available. 

gh thorium, curium, neptunium, 233U, 238Pu, and fission products have been identif

tion and/or waste products, there is no method discussed for calculating dose from

e radionuclides. There is limited information on bioassay te

ra

exposures to these radionuclides are considered in internal dose calculations. 
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ORAUT-TKBS-0011-4 (McDowell-Boyer and Little 2004, pg. 11) states the following: 

entrations, typically near the 903 Area after 1970 because concentrations 

are more often reported for samples in that area and not for samplers in other 

The da

conduc

deficien

ble. These data are reported monthly and annually, and 

Fur

hav

the flow rates through the air

sample filter did not permit concentrations of radioactive material in effluent air 

to be accurately measured. Thus, conformance with appropriate DOE and EPA 

requirements could not be demonstrated with confidence. 

Since the TBD relies heavily on environmental air monitoring data, the impact of air sample 

deficiencies on environmental release should be investigated. 

Table 4-2 lists estimated annual intakes of 239,240Pu and 241Am between 1965-

2002. The average values for 239/240

Pu in this table are based on measured 

conc

RFP areas. Therefore, the average values are inherently claimant-favorable 

when used as an average in the industrial area. 

ta collected to monitor airborne environmental releases were deficient. In 1989, DOE 

ted an assessment on the environmental conditions at the RFP. The report noted 

cies in the effluent air monitoring program (DOE 1989). 

Deficiencies in Rocky Flats Plant (RFP) effluent air monitoring program may 

adversely affect determination of radioactive materials released to the 

atmosphere.

There are deficiencies in the ambient air monitoring program for radionuclides. 

As a consequence, the accuracy of measured concentrations of plutonium in 

ambient air are questiona

are used in calculating annual radiation dose to the public to confirm dose 

calculations that are made based on radioactive effluent emission data.

The RFP does not analyze ambient air samples for americium-241. For 

preparing the dose assessments, it is assumed that the americium-241 activity is 

20 percent of the alpha-emitting plutonium activity because this is the maximum 

americium-to-plutonium ratio that may be present in RFP plutonium from the 

decay of plutonium-241. However, it is not clear that this assumption is valid 

because some processes at the RFP separate plutonium and americium and the 

latter may be released independently of the former.

The ambient air monitoring samples are not analyzed for uranium.

thermore, DOE performed an assessment on Building 559 indicating that the site continued 

ing problems with ambient air monitoring (DOE 1991). 

Releases of radionuclides from Building 559 through Building 561 were not 

measured accurately or with confidence by environmental monitoring personnel. 

The configuration of sampling equipment and
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6.3

summary, there are a number of issues related to external etry and data 

 in 

ceeded the quarterly limit, need to be 

f the integrity of the dose record. Interviews 

rt, the records associated with that 

ed

ility of sufficient data on the 241

Pu content and age of the plutonium. Specifically, 

these records would be needed for plutonium that was recycled from old warheads to 

ue will be more important from the midto late-1950s to the end of production than in the early years. 

l

he

area, and radionuclides of concern further complicate issues, making it difficult to 

5.11. Summary 

In and internal dosim

integrity that need to be disclosed and more thoroughly explored in the TBD, including the 

following: 

(1) The validity and integrity of external dose records appear to be affected by one or more 

of the following: inappropriate algorithms, calibration issues, placement of dosimeters

relation to aprons, individuals not wearing and/or improperly wearing their dosimeters,

and uncertainty with the conditions under which zero were assigned. In regard to the 

latter issue, the integrity of records containing zeros may remain open to question until 

the practice of recording zeros when badges were not returned, along with personnel 

practices such as disciplining workers who ex

carefully investigated for their implications o

with the GAO personnel who wrote the 1994 repo

report, as well as the people interviewed for its preparation could serve as a valuable 

guide to gauge the significance and extent of this problem and its practical importance for

dose reconstruction. The existence of blank records in the post-1964 period also raise

questions about the integrity of data recording practices that need to be explicitly 

addressed and resolved. 

(2) The removal of 241Am from aged Pu when it was reworked to make new pits after

obsolete warheads were retired bears further study because the amount of 241Am creat

by the decay of 241Pu depends on the initial 241

Pu concentration. That concentration 

decreases once a warhead is returned after some years in service for dismantlement and 

reuse of Pu in new designs. The validity of Pu intake estimates depends on the 

availab

fabricate new ones. SC&A expects that this iss

(3) With internal dosimetry, there were individuals not monitored, delinquencies in submitta

of bioassay samples, and questionable effectiveness in bioassay techniques to detect all 

chemical forms of radionuclides present at the facility. 

(4) The limited information on job titles, skills and tasks, work locations, time spent in t

differentiate between radiological and non-radiological workers. 

(5) The source data for calculation of airborne environmental dose is questionable, based on 

effectiveness of environmental air sampling. 
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sment 

e

ctives defined by the Advisory Board for ascertaining adequacy; 

rofile 

t

plete 

tory.

tion of these issues is provided elsewhere 

n.

esses and personnel monitoring data. 

equate

6.0 OVERALL ADEQUACY OF THE SITE PROFILE AS A BASIS FOR 

DOSE RECONSTRUCTION 

The SC&A procedures call for both a “vertical” assessment of a site profile for purposes of 

evaluating specific issues of adequacy and completeness, as well as a “horizontal” asses

pertaining to how the profile satisfies its intended purpose and scope. This section addresses th

latter objective in a summary manner by evaluation of (1) how, and to what extent, the site 

profile satisfies the five obje

(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 p

be evaluated against five measures of adequacy—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 RFP represents an

adequate accounting of the “core” plutonium and uranium and dosimetry history of the plant, bu

falls short in fully characterizing underlying issues that are fundamental to guiding dose 

reconstruction. In some cases, the issues will impact other site profiles. Many of the issues 

involve lack of sufficient conservatism in key assumptions or estimation approaches, incom

site data or incomplete analyses of these data, or incomplete reflection of the operational his

Section 6.0 summarizes the key issues. Detailed evalua

in the report.

6.1.1 Objective 1: Completeness of Data Sources 

The breadth of data sources used as a basis for the RFP site profile is evident in the 415 reports 

available in the Site Profile Research Database. One hundred and nine (109) of these reports 

were cited in the site profile references, while others served to provide confirmatory informatio

The ORAU team included health physics personnel with long histories at RFP who have 

extensive knowledge of key dosimetry historical proc

However, a number of historical issues and discrepancies cast doubt on the validity, and in some

instances, the integrity of dose records being relied upon for EEIOCPA dose reconstruction, ones 

which are not adequately reflected in the TBDs that make up the site profile. Evidence exists 

that there is a potential for missed occupational external, internal, and environmental dose due to 

the incompleteness and questionable quality of data being used in dose reconstruction. These 

missed doses may have occurred as a result of incomplete monitoring records and inad

monitoring techniques. There may also be a data integrity issue associated with the external dose 

record. 
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e 

n,

is

opes other than uranium,

lutonium, and americium. 

s

riety of

dionuclides ( Sr, Cs, uranium, Pu, Pu, Tl, Pb, Pb, and Np) have been 

i quately address potential dose from recycled 

ranium, highly enriched uranium, and the processing of 233U. 

e TBD. 

xposure data on radiological 

cidents. Workers have often complained of poor record keeping and fabrication of records, 

ord keeping practices that might indicate problems with the integrity of

dividual worker data in some settings and periods.

ied

or

wt the

e

In general, the site description TBD (Flack and Meyer 2004) provides an accurate and reasonabl

characterization of the history and operational activities at the RFP for the period of productio

but falls short in doing this for the post-operational decontamination and decommissioning

activities. The major functions of many of the production and support buildings were outlined, 

as well as the major process changes that occurred throughout the years. In addition, 

accident/incidents, such as fires, were documented in chronological order. The TBD

incomplete in its review of the historic dose contribution of radioisot

p

The TBD Site Description provides useful information regarding radionuclides handled in 

different buildings. But, it does not provide a comprehensive list of radionuclides and quantitie

of radionuclides to which workers may have been exposed. For example, a va

89/90 134/137 238 239 208 212 214 237 ra

identif ed in the soil. The TBD does not ade

u

The TBDs do not adequately cover the dose potential from decontamination and 

decommissioning activities that have been ongoing since the end of the production period in 

1989, and processing of special materials such as recycled uranium and special project 

radionuclides. The occupational environmental dose TBD (McDowell-Boyer and Little 2004) 

provides limited direction to dose reconstructors on the process and assumptions that should be

used to calculate internal dose. The TBD provides limited guidance for the assessment of missed

dose for unmonitored workers. NIOSH has indicated that this guidance is provided in 

supplementary procedures and technical information bulletins (Attachment 4), which is not 

inherently obvious from the information in th

For monitored workers, ORAU and NIOSH rely primarily on the worker dose record and the 

Computer Assisted Telephone Interview (CATI) for potential e

in

which have been confirmed in Congressional and Tiger Team investigations. The Rocky Flats

Site Profile, as it now stands, does not have a dedicated database or assessment for off-normal

exposures and incidents that can guide dose reconstructors as to missed incidents (and hence 

additional doses) or rec

in

The occupational medical TBD (Furman and Lopez 2004) does not show that the RFP appl

dose minimization principles to reduce medical exposures. The document also does not assess

consider the likely exposure to workers who are referred to off-site medical facilities for follo

up. The TBD states that review of selected medical records and files did not reasonably show or

match expected x-ray exam frequency, and type of exam. Little evidence exists to documen

number of x-ray exams provided to the average worker or for special exposure needs.

ORAUT-TKBS-0011-6 (Langsted 2004) describes the default assumptions for occupational 

external dose at RFP. ORAUT-OTIB-0027, Supplemental External Dose Information for Rocky

Flats Plant (Smith 2005), provides additional information on the assumptions to be used in dos
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imetry 

om the dosimetry report review. The site profile refers the 

ose reconstructor to the Rocky Flats work history file, the Neutron Dose Reconstruction 

o

number of deficiencies were identified with the internal TBD (Falk 2004). NIOSH’s use of 

n

es in 

tify additional particle size studies conducted at RFP; 

owever, we were not able to obtain copies. These data should be taken into consideration when 

t

um,

nt

t was

t records. The Neutron Dosimetry Reconstruction Project Protocol 

DRP) was designed to remedy these deficiencies by recounting original tracks and providing a 

ch

reconstruction. The assumptions were derived from historical records relating to the dos

program and observations made fr

d

Protocol (NDRP) file, and a job exposure matrix put together by the University of Colorad

Health Sciences Center and the CDPHE for additional information. The job exposure matrix 

data is unavailable. The NDRP information has only recently become available to 

NIOSH/ORAU and has not yet been integrated into the site profile. 

The TBDs (Little and Meyer 2004, Flack and Meyer 2004, and Langsted 2004) do not mention 

any use of industrial x-ray units, large radioisotope sources, or neutron generators for R&D or 

non-destructive testing (NDT), or any associated radiation exposure or monitoring. However, 

these units (some in the MeV energy range) were apparently used at the RFP. 

6.1.2 Objective 2: Technical Accuracy 

A

median MDA (minimum detectable activity) values for plutonium and americium appear unduly 

low and likely to yield body burdens or organ doses that would be non-conservative, given the 

uncertainties involved. The default particle size for dose reconstruction is 5 μm AMAD. Man

and Kirchner (1967) measured a particle size of 0.3 μm mass median diameter (1-2 μm AMAD) 

during a plutonium fire. Particle sizes and distributions are not available for all work areas or

incidents; however, given the production process, the potential exists for smaller particle siz

work areas. SC&A has been able to iden

h

evalua ing the default particle size. 

The solubility and particle size should be assigned based on the most claimant-favorable value 

for the particular type of cancer, 

The TBD does not provide adequate guidance for dose reconstruction associated with triti

neptunium, thorium, curium, noble gases, and fission products. Ingestion exposure pathways 

were not considered in the TBD in the dose assessment process, especially for cancers of the

gastrointestinal tract. 

The external dosimetry sections of the TBDs and supporting OTIBs, while providing pertine

historical data and technology information on dosimetry systems and records, fall short in 

addressing potential missed and unmonitored dose, particularly in the early years of operation 

(1950s through mid-1960s). The use of neutron track plates and the uncertainties in neutron 

track counting with these and NTA film indicates there may be important missed dose tha

not entered into claiman

(N

corrected estimated individual neutron dose. However, the recently issued NDRP report does 

not cover non-Pu workers, nor is it applicable to unmonitored or non-neutron workers. There is 

a resultant need to use neutron/photon ratios and film/TLD comparisons to correctly determine 

past neutron doses. The TBD (Langsted 2004) only briefly addressed these two issues, whi
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nal medical dose TBD does not adequately address the contribution of 

istoric radiation exposure from occupationally necessitated medical x-ray exposure of workers 

NIOSH guidelines (OCAS 2002a) to 

include only one pre-employment and a chest x-ray annually, it underestimates dose in that it 

y

ssumptions derived for other site profiles (e.g., Savannah River occupational medical TBD) 

ed

active material in soil excluded these other 

reas. The ambient air monitoring data is of questionable validity for use in assessing onsite 

g

y, issues include established problems 

ith algorithms, calibrations, dosimeter placement (including in relation to lead aprons), 

ro or

n,

ittals, 

P

limitations. 

The issue of unmonitored workers, particularly from the early 1950s to the early 1960s, is not 

developed in the TBDs from the standpoint of the extent of the problem and how dose estimation 

would be handled, particularly the application of co-worker doses. Assignment of dosimeters to 

workers using the “10% of the radiation protection guide” policy was apparently based on 

could be important in generating correction factors for under-monitored workers or for 

monitored-worker missed dose. 

The RFP occupatio

h

at Rocky Flats. By narrowly interpreting pre-established

excludes from consideration other special work-required x-ray procedures, e.g., x-rays required 

for respiratory protection certification, and special exams for asbestosis and beryllium workers 

(which often occurred at frequent intervals and not as part of an annual physical). Also omitted 

from consideration are x-rays resulting from work-related injuries. Consequently, by narrowl

interpreting what exposure sources should be included and focusing on the relevancy of 

a

rather than on the actual adequacy and accuracy of existing RFP data, the site profile falls 

considerably short in characterizing this source of occupational exposure. 

The RFP environmental TBD excluded internal dose as a result of soil resuspension and focus

its efforts on only one soil contamination area without justification. Soil contamination has been

found at many areas on the Rocky Flats site, including in the East Spray Fields and the buffer 

zone. There is no explanation of why dose from radio

a

worker environmental dose. Annual inhalation intake values developed from atmospheric 

modeling appear to be underestimated. 

6.1.3 Objective 3: Adequacy of Data 

A number of historical issues and discrepancies cast doubt on the validity of dose records bein

relied upon for EEOICPA dose reconstruction, ones which are not adequately reflected in the 

TBDs that make up the site profile. For external dosimetr

w

individuals not wearing or improperly wearing dosimeters, and conditions under which ze

null doses were assigned. In some cases, it is clear that zeros were recorded in lieu of more 

legitimate dose estimates. For internal dosimetry, particularly for the first 20 years of operatio

missing dose likely resulted from exposures below the Minimum Detectable Activity (MDA) for 

plutonium that were recorded as zero or background, inadequate monitoring (urinalysis and 

fecal) following incidental plutonium exposures, delinquencies in bioassay sample subm

and questionable capability to detect all chemical forms of radionuclides present. Although the

TBDs incorporate methodologies for assigning missed dose, it is not clear that these bound RF

exposure conditions and compensate for errors introduced by calibration and monitoring 

technology
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n extensive comparison was performed by SC&A to compare and contrast the methodologies 

s

port.

of 1.0 is applied for environmental dose. This value is claimant favorable. 

he profile also introduces an inhalation dose factor from ICRP 72, Age-dependent Doses to 

nd

tion

ultiple receptor points onsite. The RFP site profile used onsite environmental air sampling data 

ot

quately monitored 

onstitutes a different set of individuals. Those that were not monitored at all were likely to be 

ore clearly stated to exclude inadequately monitored workers who actually received 

xposure.

administrative decisions, not necessarily on actual survey results, and supported by site expert

interviews. Although early neutron monitoring data can be supplemented by the NDRP study,

this report excludes non-plutonium workers, workers not monitored with NTA film, and nonneutron workers. 

6.1.4 Objective 4: Consistency Among Site Profiles 

A

used in the RFP and other site profiles reviewed to date to determine external, internal, medical, 

and environmental dose. These comparisons focus on the methodologies and assumption

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 5 to this re

The RFP site profile has introduced a few new concepts not previously used in other site profiles.

A respirable fraction

T

Members of the Public from Intake of Radionuclides: Part 5: Compilation of Ingestion a

Inhalation Dose Coefficients (ICRP 1996) for select radionuclides. An inhalation dose frac

has not been included in other site profiles reviewed to date. It is unclear why this factor would

apply to RFP, but not to other DOE and AWE sites. 

Site profiles such as Hanford, SRS, and Idaho National Laboratory (INL) have included data for 

m

and atmospheric dispersion model estimates based on public dose assessments to determine a 

single intake quantity per year. Presumably, the assumption here is that the onsite air monitors 

will provide cumulative release data from all operations of the plant. The site profile has n

demonstrated that this single set of intake data represents the highest potential environmental 

exposure onsite. A similar analysis at multiple receptor points is warranted to determine whether 

this single intake quality is bounding for all areas onsite. 

Environmental occupational dose is typically assigned to those individuals who were not 

monitored. In the case of the RFP site profile, environmental dose is applied when “a worker 

was not monitored adequately to develop a reliable individual dose.” A worker who was not 

monitored for either external or internal dose versus a worker that was not ade

c

in positions where radiation exposure was not an issue. Workers who were monitored 

inadequately can’t be compared to those who may or may not have been exposed to radiation. 

The terminology used for the application of occupation environmental dose is confusing, and 

should be m

e
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d Bihl 2005).

lthough RFP is not a reactor facility, the 28-radionuclide intake scenario for facilities with 

pothetical intake for

a reactor versus non-reactor facility. It is important to note that the internal dosimetry site profile 

,

e of less than 

ference man or woman be normalized to these values. A consistent method of addressing 

le collection, the approach in the Hanford site profile is more 

laimant favorable. 

t

rocessing area) may 

tons present in plutonium

s

ced, this was no longer necessary. RFP also recognized the inadequacies in the twon

tion factors out of the data in order to separate the 

ave assigned missed tritium dose. The internal 

pr issed tritium dose. A similar issue occurs 

ear when missed neutron dose should be 

CFR

ult 5μm 

The maximizing approach for missed internal dose to unmonitored workers is determined by 

using ORAUT-OTIB-0002 (Rollins 2004) or ORAUT-OTIB-0018 (Brackett an

A

reactors was used. NIOSH/ORAU has provided no rationale for using the hy

does not mention the use of the above procedures for determining a maximum dose to 

unmonitored workers. There is also no explanation for when to use which TIB method. 

In the RFP site profile, it was assumed that the sample volume represented a 24-hour collection

and no corrections were made to the volume whether they represented an actual 24-hour 

collection or not. In the Hanford site profile, it is recommended that a urine volum

re

sample volume is needed. Ultimately, determining individual 24-hour sample volumes is ideal.

In the absence of true 24-hour samp

c

There has been no consideration of ingestion dose from workers at the RFP facility. The 

Bethlehem Steel and Mallinckrodt Chemical Works site profiles have included dose from

ingestion of radioactive material. Although there may have been more engineering controls a

RFP, failure of these engineering controls, internal contamination of respirators, incidents, and 

allowing food and beverages in specific areas of the plant (e.g., uranium p

have led to ingestion of radioactive material. 

Two-element dosimeters did not effectively account for low energy pho

facilities. This was recognized by the sites and adjustments were made. For example, SRS used 

a special x-ray calibration curve for workers involved with handling plutonium. Hanford 

included 20% of the open window dose in the penetrating dose. Once better designed dosimeter

were introdu

and three-element dosimeters through 1970 when the TLD was implemented. For the twoelement dosimeter, they added 50% of the open window dose to the penetrating dose. For the 

three-element dosimeter, they added 35% of the open window dose to the penetrating dose. I

the RFP TIB, ORAU has backed these correc

photons into energy categories. This was not done with data from Hanford and Savannah River 

Site. The difference in the RFP approach has not been adequately explained. 

Some sites such as the Savannah River Site h

TBD ovides no guidance on when and if to assign m

with the assignment of missed neutron dose. It is uncl

assigned. 

6.1.5 Objective 5: Regulatory Compliance 

With one exception, NIOSH has complied with the hierarchy of data required under 42

Part 82 and its implementation guides. 42 CFR Part 82 recommends the use of the defa
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e 

e following seven objectives to guide its

view of the RFP Site Profile TBDs to determine whether it meets these criteria: 

select

not affect the outcome. 

potential exposures and

,

time and employment locations. 

ect hich the claimant is 

n

6.2 A

In the DR process, the assignment of ISO or ROT instead of AP geometry may not reflect the 

true i

for o

always b

investiga

particle size only in cases where there is no information on particle sizes. This is not the cas

with the RFP facility, as stated in ORAUT-TKBS-0014-5 (Falk 2004, pg. 9). 

6.2 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 th

re

Objective 1 − Determine the degree to which procedures support a process that is expeditious 

and timely for dose reconstruction. 

Objective 2 − Determine whether procedures provide adequate guidance to be efficient in

instances where a more detailed approach to dose reconstruction would

Objective 3 − Assess the extent to which procedures account for all

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

Obj ive 5 − Evaluate procedures with regard to fairness and the extent to w

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 distributio

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.

.1 mbiguous Dose Reconstruction Direction

rad ation dose to some workers; such as 100% rotational for site support personnel or 50% 

supp rt personnel as given in the TBD (Langsted 2004, Table 6-5, pg. 23). This may not 

e claimant favorable in some cases (i.e., for anterior-seated cancers). Further 

tion should be done concerning this issue and revised DR instructions issued.
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In t

radiation

For example, the TBD (Langsted 2004, Section 6.6.4.1, pg. 27), mentions photon angular 

dep ry

response esponse

decreased as the angle increased. Page 40, Section 6.8.4.1, addresses beta angular dependence 

and e

for TLD

his TBD, and other TBDs as well, the issue of angular dependence for different types of 

and different dosimetry system used through the years is not sufficiently addressed. 

endence, but seems somewhat vague about the issue. On page 27 it states that the dosimet

increases with angle, but on page 28 it states that the Panasonic system r

stat s that the assembled badge displays severe angular dependence to beta exposure and that 

badges it might record only 36% to 59% of the true beta dose at +30 degrees (this

underestimating the true b

could

result in eta dose, especially to the extremities where the worker’s

normal movements would not tend to average it out). 

The TBD is

supposed TOTIB-0002 (Rollins 2004) or ORAUT-OTIB-0018 (Brackett and Bihl 2005). There is no 

dire dose is 

absent fr

radionuc

6.2 I

SC&A n

included e inconsistencies and

editorial errors should be corrected. 

6.3 U

A number of issues were identified that are common in the RFP and other site profiles reviewed 

to d idual

site profi e interpretation of existing standards (e.g., oro-nasal 

reathing), 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 RFP Site Profile review that, in SC&A’s view, 

represent transcendent issues that need to be considered by NIOSH as unresolved policy or 

generic technical issues.

(1) Direction on the applicability of the TBD and/or TIBs to individual dose reconstructions 

is absent. 

(2) Mobility of the work force between different areas of the site should be addressed. Site 

expert testimony that many workers moved from one plant to the next is a complicating 

factor. Establishment of an accurate worker history is crucial in such cases. This will 

be especially difficult for family member claimants. 

addressed internal dose to unmonitored workers in a limited manner. The reviewer

to infer from the TBD that maximizing internal doses are calculated by using ORAU

ction on when which TIB should be used. Also, the assignment of missed tritium

om the TBD. There is no guidance provided on how to assess internal dose from

lides other than plutonium, uranium, tritium, and americium.

.2 nconsistencies and Editorial Errors in the Site Profiles 

oted some inconsistencies and editorial errors in the site profile TBDs which are 

in Attachment 7. In future revisions to the RFP TBDs, thes

NRESOLVED POLICY OR GENERIC TECHNICAL ISSUES 

ate and, in some cases, represent potential generic policy issues that transcend any indiv

le. These issues may involve th

b
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(3) Statistical techniques used individual workers should be

further considered and substantiated. 

tributory exposure source.

uence

dose contributions need to be considered (most notably) for solubility, oro-nasal 

eds to be performed regarding how “frequent or routine incidents” should be 

addressed, given the possibility that such “spike” exposures may be often missed by 

ucted.

may have potential exposure 

while working on or visiting a facility should be ascertained. 

&D operations, which extend back many years. 

Decommissioning and decontamination operations often required working in unknown 

ccupational medical exposures remains incomplete. NIOSH 

needs to reconsider the definition to include all forms of radiation medical exposure, to 

(10) Dose contributions from chemical forms of radionuclides that are not adequately 

in the application of the data to

(4) Dose from impurities and/or daughter products in radioactive material received and 

processed at sites should be assessed as a con

(5) The significance of various exposure pathways and the assumptions made that infl

breathing, and ingestion. 

(6) Analysis ne

routine monitoring as a function of how often and in what manner it was cond

(7) Availability of monitoring records for “transient or outside workers,” e.g., 

subcontractors, construction workers, and visitors who

(8) Dose to decontamination and decommissioning workers should be assessed. Many 

facilities have large-scale D

situations, which may provide unique exposure situations. 

(9) Dose reconstruction for o

ensure its considerations are claimant favorable. 

addressed in the ICRP models. 
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R Part 82, 2002. Title 42, Public Health, Part 82, Methods for Radiation Dose 

econstruction Under the Energy Employees Occupational Illness Compensation Program Act 

ames Ruttenber, September, 12, 2005. 

cal

imeline of External Dosimetry Equipment Used at Rocky Flats Since 1952, SCB-031-02, 

e, B., 2005. Personal Communication with Bob Bistline. September 19, 2005. 

r

ugust 9, 2005.

f Findings.

hemRisk (ChemRisk, a division of McLaren/Hart) 1992. Reconstruction of Historical Rocky 

I:

hemRisk (ChemRisk, a division of McLaren/Hart) 1994a. Estimating Historical

hemRisk (ChemRisk, a division of McLaren/Hart) 1994b. Dose Assessment for 

Surveillance Program at Rocky Flats, Health Physics

7.0 REFERENCES 

42 CF

R

of 2000, Code of Federal Regulations, May 2, 2002. 

Alvarez, R., 2005. Personal Communication with J

Baker, S., 2002. Memorandum from Steve Baker to the External Dosimetry File, Histori

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from Occupationally Related Diagnostic X-ray Procedures, ORAUT-OTIB-0006, Rev. 03, O

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Occupational External Dosimetry, ORAUT-TKBS-0014-6, Rev. 00, Oak Ridge Associated 

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Tennessee. January 20, 2004. 

Lincoln, T.A. and

R

Lipsky, J.S., 1989, Application and Affidavit for Search W

D

L

ORAUT-TKBS-0011-1, Rev. 00, Oak Ridge Associated Universities, Oak Ridge, Te

A

Mann, J.R., 1964. Letter to J.B. Owen, H

D

Mann, J.R., and Boss, M.R., 1963. Neutron Dosimetry Problems in

P
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auro, John. 2005. Personal Communication with Paul G. Orlando, Chief, Bureau of 

ell-Boyer, L. and Little, C., 2004. Technical Basis for the Rocky Flats Plant – 

ccupational Environmental Dose, ORAUT-TKBS-0011-4, Oak Ridge Associated Universities, 

atural Resources Defense Council (NRDC), Archive of Nuclear Data, Figure of U.S. Nuclear 

ction for Energies Up to 50 MeV (Equipment Design, 

erformance, and Use), Report No. 102, NCRP, Bethesda, Maryland. 

orton, M.J, Hartman, B.M., Fimberg, K.R., and Murtha, P.J., 1992. Plaintiff’s Sentencing 

ion

Prevention, U.S. Department of Health 

nd Human Services, Washington, DC. 

h

assmore, C.N., 1995. Single Sided Lead Apron Study, Pantex Plant, May 2, 1995. 

on the Five Major

cidents at Rocky Flats. July 11, 1986.

m and Americium in

issues from a Human Autopsy Case. J. Radiol. Prot. 9, 159-164.

Mann, J. R. and Kirchner, R. A., 1967. Evaluation of Lung Burden Following Acute Inhalation

Exposure to Highly Insoluble PuO2, Health Physics, Volume 13, pp. 887-882. 

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Radiological Health, New Jersey Department of Environmental Protection. October 11, 2005. 

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Oak Ridge, Tennessee. June 29, 2004. 

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Stockpile, 1945-2002, Last revised November 2, 2002, 

http://www.nrdc.org/nuclear/nudb/dafig9.asp 

NCRP (National Council on Radiation Protection and Measurements) 1989. Medical X-ray, 

Electron Beam, and Gamma-Ray Prote

P

N

Memorandum, United States of America (Plaintiff) vs. Rockwell International Corporation 

(Defendant), United States District Court for the District of Colorado. 

OCAS (Office of Compensation Analysis and Support) 2002a. External Dose Reconstruct

Implementation Guidelines, Revision 1, OCAS-IG-001, National Institute for Occupational 

Safety and Health, Centers for Disease Control and

a

OCAS (Office of Compensation Analysis and Support) 2002b. Internal Dose Reconstruction 

Implementation Guide, OCAS-IG-002, National Institute for Occupational Safety and Health, 

Cincinnati, Ohio. 

ORAU (Oak Ridge Associated Universities) 2003. Technical Basis Document for the Savanna

River Site, ORAUT-TKBS-0003, Oak Ridge, Tennessee. 

Owen, J. B., 1964. Handling and Processing Dosimeter Film at Rocky Flats Division.

P

Plott, Carmine, 1986. Memorandum to Larry Goldren. Subject: Files

in

Popplewell, D.S. and Ham, G.J., 1989. Distribution of Plutoniu

T

Putzier, E.A., 1982. The Last 30 Years at Rocky Flats Plant.
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ort – Year 1984, Radiation Dosimetry Section of the Environmental & 

ccupational Health Branch, Rocky Flats Plant, Golden, Colorado, December 1985. 

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0,

ect,

ockwell International 1986. 707 Building Safety Audit, Rocky Flats Plant, Golden, Colorado. 

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tional, Rocky Flats Plant, Golden, Colorado, June 1987. 

ssee, May 7, 2004.

, RAC

Neeses, South

arolina, September 1999. 

ental Data for Historical Public Exposure Studies on Rocky 

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lats, Los Alamos National Laboratory, LA-UR-05-3247, May 2005. 

uttenber, A.J., et al. 1996. Health Studies On The Rocky Flats Plant, Prepared by Colorado 

epartment of Public Health and Environment, October 1996. 

uttenber, A.J., et al. 2003. Improving Estimates of Exposures for Epidemiologic Studies of 

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RFP (Rocky Flats Plant) 1985. Radiation Exposure to Rockwell Employees at the Rocky Flats 

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RFP (Rocky Flats Plant) 1988. Personnel Protection Requirements for Plutonium Controlled 

Areas, HSE

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RFP (Rocky Flats Plant) 1992. External Radiation Dosimetry, 1-16200-HSP-18.07, Revision

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Rockwell International 1987. 1986 Annual Report Radiation Exposure to Rockwell Emp

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Rollins, G., 2004. Technical Information Bulletin, Maximum Internal Dose Estimates for 

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Universities, Oak Ridge, Tenne

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T.F., 1999. Evaluation of Environm

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R

F

R

D

R
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S ,

March 18, 2004. 

nt Review of the NIOSH Site Profile Review for Mallinckrodt 

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vannah River Site To Be Used for 

-

E-RFP-1999-Final, Radiological Assessment 

ium in Airborne Effluents at Rocky 

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al

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Scalsky, E.D., 2004. Technical Basis Document for the Sa

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Ridge Associated Universities, Oak Ridge, Tennessee.

Smith, M.H., 2005. Supplementary External Dose Information for Rocky Flats Plant, ORAUT

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Voillequé, P.G., 1999a. Estimated Airborne Releases of Plutonium During the 1957 Fire in 

Building 71, RAC Report 10-CDPHE-RFP-1999-Final, Radiological Assessment Corporation, 

Neeses, South Carolina.

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Building 776/777, RAC Report 9 – CDPH

Corporation, Neeses, South Carolina. 

Voillequé, P.G., 1999c. Review of Routine Releases of Pluton

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South Carolina, August 1999.

Weber, J.M., A.S. Rood, H.R. Meyer, and J.E. Till, 2005. Development of the Rocky Flats Plant 

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Radiological Assessments Corporation, Neeses, South Carolina, August 1999. 

Widner, T., 2005. Technical Basis for the Los Alamos National Laboratory – Occupation

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Ridge, Tennessee. May 10, 2005.

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of the Hanford Personnel Dosimeter Program from 1944 through 1989. PNL-7447. Pacific 

Northwest Laboratory, Richland, W
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ATTA ED

DURING THE REVIEW 

Technical Basis Documents 

RAUT-TKBS-0011-1, Technical Basis Document for the Rocky Flats Plant - Introduction 

ev. 00 (Little and Meyer 2004) 

RAUT-TKBS-0011-2, Technical Basis Document for the Rocky Flats Plant – Site Description 

Rev. 00 (Flack and Meyer 2004) 

ORAUT-TKBS-0011-3, Technical Basis Document for the Rocky Flats Plant – Occupational 

Medical Dose Rev. 00 (Furman and Lopez 2004) 

ORAUT-TKBS-0011-4, Technical Basis Document for the Rocky Flats Plant – Occupational 

Environmental Dose Rev. 00 (McDowell-Boyer and Little 2004) 

ORAUT-TKBS-0011-5, Technical Basis Document for the Rocky Flats Plant – Occupational 

Internal Dose Rev. 00 (Falk 2004)

ORAUT-TKBS-0011-6, Technical Basis Document for the Rocky Flats Plant – Occupational 

External Dosimetry Rev. 00 (Langsted 2004) 

Technical Support Documents 

ORAUT-OTIB-0002, Rev. 01 PC-2, (2004), Technical Information Bulletin, Maximum Internal 

Dose Estimates for Certain DOE Complex Claims, Oak Ridge Associated Universities, Oak 

Ridge, Tennessee, May 7, 2004 (Rollins 2004) 

ORAUT-OTIB-0006, (2005), Technical Information Bulletin: Dose Reconstruction from 

Occupationally Related Diagnostic X-ray Procedures, Rev. 03, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee. August 2, 2005. (Kathren and Shockley 2005) 

ORAUT-OTIB-0018, Rev. 01, (2005), Internal Dose Overestimates for Facilities with Air 

Sampling Programs, Oak Ridge Associated Universities, Oak Ridge, Tennessee, August 9, 2005, 

(Brackett & Bihl 2005) 

ORAUT-OTIB-0027, (2005), Supplementary External Dose Information for Rocky Flats Plant, 

Oak Ridge Associated Universities, Oak Ridge, Tennessee, May 15, 2005.(Smith 2005) 

ORAUT-PROC-0003, Internal Dose Reconstruction, Rev. 00, Oak Ridge Associated 

Universities, Oak Ridge, Tennessee, May 1, 2003 (Brackett 2003) 

CHMENT 1: NIOSH TECHNICAL DOCUMENTS CONSIDER

O

R

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ED>

ATTACHMENT 2: SITE EXPERT INTERVIEW SUMMARY 

<RESERV
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ATTACHMENT 3: KEY QUESTIONS FOR NIOSH/ORAU REGARDING 

IT

1. for external penetrating radiation? How 

2. 

before the 1964 transition to a

ould

n

al projects such as processing/fabricating of 

233

233

6. at

ats?

d

9. 

ion products? The TBD appears to have no 

10.

pes from fires at Rocky 

nal and external exposure? 

f

12.

SITE PROFILE DOCUMENTS 

S E DESCRIPTION (ORAUT-TKBS-0014-2)

What were the criteria for monitoring workers

did these criteria evolve over time?

What percentage of Rocky Flats Plant (RFP) workers were measured for external

radiation over time, including neutrons and photons?

3. How are unmonitored workers addressed, particularly

combined security/dosimetry badge? For example, guards were apparently not routinely 

monitored, although they were often stationed adjacent to production areas; how w

their exposures be characterized?

4. What data did NIOSH/ ORAU obtain regarding the degree and extent of contaminatio

of specific processes that resulted from speci

U and other radioisotopes?

5. How much U was handled at Rocky Flats over what periods of time?

What were the levels of 232U contaminants present in the 233

U processed and handled

Rocky Fl

7. Why don’t the TBDs account for recycled uranium of various assays that were processe

at Rocky Flats?

8. What processes at Rocky Flats would have been likely to concentrate contaminants 

contained in recycled uranium?

Were workers who handled and processed recycled uranium measured for exposure to 

contaminants such as 239

Pu, 237Np, and fiss

information about this issue of concern. 

What radiological controls initially were established at Rocky Flats and how did they 

subsequently evolve over time?

11. How many workers were exposed unexpectedly to radioisoto

Flats? Of this group, how many were not monitored for inter

Of those monitored, how many (on a fire-by-fire basis) received doses in excess o

prescribed limits? 

How many workers entered airborne radioactivity areas without respirators and how did 

they assess or estimate these workers’ inhalation dose?
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4. What criteria and guidance were used to verify criticality safety relative to drains and

5. What parameter controls (i.e. mass, enrichment, density, concentration, geometry, 

8. Were there any incidents involving the furnace operations with respect to ignition of 

19. How did NIOSH determine which incidents were to be included in the site profile?

ny

contacts for these organizations? 

nal

ments and job titles. Has the matrix 

been obtained by NIOSH for use in DRs?

23. l

ith significant external 

and internal dose potential in the early days when RFP workers performed a lot more 

OCCU

1. The above listed document is dated February 2004. It is noted that a lot of cross 

currently doing

to provide a revision to this site technical basis document? Is NIOSH currently revising 

13. How many workers required chelation therapy during the history of the operation of 

Rocky Flats?

1

overflows that may have evolved over time?

1

interaction, reflection, moderation, and neutron absorption) were applied to prevent 

criticalities at Rocky Flats and how did they evolve?

16. Did Rocky Flats handle or process recycled thorium?

17. What assessments of criticality safety and radiological control at Rocky Flats did 

ORAU/NIOSH review in the preparation of the TBD?

1

combustible gasses, flame flashbacks, etc.? 

20. What records retrieval efforts were made in support of the site profile review? Were a

classification issues encountered during site expert interviews and records retrieval 

efforts?

21. What worker organizations were contacted to obtain worker input for the site profile? 

Who are the

22. A “job exposure matrix” was developed as part of a DOE-funded study performed by the 

University of Colorado Health Sciences Center and the Colorado Department of Public

Health and Environment (and cited in ORAUT-TKBS-0011-6), which matches exter

dosimetry results with year-by-year building assign

Page 7, Section 2.4.3.1, states that DU is not generally considered a significant externa

or internal hazard. Why isn’t uranium a significant radionuclide w

operations with uranium (enriched, natural, and depleted?

PATIONAL MEDICAL DOSE (ORAUT-TKBS-0014-3) 

referencing is relied upon between various site profiles. What is NIOSH

any of the TBD documents for other sites that would require subsequent changes to this 

document?
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2. ut the existence or absence of medical protocols used at 

RFP to determine the need for radiological exams during his tenure at RFP or prior to his 

3. Were x-rays used routinely as part of medical certification for respirator users, food 

4. tion exposure? Does it include 

evaluation of foreign material resulting from machining and milling accidents?

5. e?

6. The TBD suggests all radiography occurred in the diagnostic range of (30-250 kVp). Is 

7. ysical evidence (by medical record 

or protocol documents) that show workers were not exposed to voluntary chest x-rays, 

8. 

rovided in the technical document?

e technical

10. Are there any records or protocols to demonstrate that radiological processing equipment 

was properly maintained and serviced to reduce and minimize worker exposure? What 

11. Did the RFP site acquire and use portable x-ray units? Can NIOSH better describe the 

utilization of photofluorography (PFG) units at RFP? What was the make and model?

d?

ployer and as a condition of employment? An example 

would be follow-up exposure to evaluate the presence of chronic beryllium disease.

OCCU

How does the analysis of releases of radionuclides from RFP differ between the Phase I 

operational period (1952 to 1994) and the Phase II post-1992 phase when production 

What does Dr. Furman know abo

time at RFP? 

handlers, nurses, etc? 

Define what is meant by occupational medical radia

Were workers required to have radiography exams as part of a return to work clearanc

NIOSH aware of the acquisition of any x-ray units that were installed at RFP that 

exceeded 250 kVp? 

Section 3.2 discusses exam frequency. Is there any ph

even when policy did not require them?

Is there any physical evidence to show that special surveillance exams for beryllium,

asbestosis, and respiratory protection were always scheduled as a part of the annual 

physical which is an assumption p

9. Are there any records or studies to document a low retake rate alluded to in th

document?

records exist for the purchase and maintenance of all x-ray equipment, radiographic 

processors and supplies for the years 1953–1989? 

What was the period of use? How many exams were performe

12. Many radiographic procedures are performed offsite on a medical referral basis. Does

NIOSH have evidence to determine the amount of offsite radiographs performed on 

workers at the request of the em

PATIONAL ENVIRONMENTAL DOSE (ORAUT-TBKS-0014-4) 

1. 
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2. II? Why is there a 1-year 

overlap between the two phases?

Describe in more detail how atmospheric dispersion modeling results, from the Phase II 

4

Explain why it might not have been better to define the phases by the types and quantity 

l status at RFP?

and models dealt with particles < 30μm AED? Would NIOSH describe the rationale for 

6. nvironmental TBD ignore the potential for ingestion of 

contaminated soils and other windblown particulates that are larger than normally 

y

Why is the resuspension of Pu and Am deposited on soil not included in any 

n in the

8. 

activities ceased and releases were more likely to occur as result of past contamination 

and decontamination activities?

Please provide a more clear definition of Phase I and Phase

3. 

study, were used to estimate air concentrations and intakes during the 1953 to 196

period when air sampling data were not available?

4. 

of data available rather than basing and restricting the modeling assumptions and the 

results on the operationa

5. What is the relationship between the 15μm aerodynamic equivalent diameter (AED) and 

the 5μm activity median aerodynamic diameter (AMAD)? Why is it that the predictions

applying these different parameters for particle size? 

Why does the occupational e

respirable particles that are traditionally included in dose assessments? What is the basis

for not including inadvertent ingestion of radioactively contaminated soil and other finel

dispersed radioactive materials?

7. 

calculations? Has NIOSH taken into account that the continuous resuspension of 

deposition during previous years would contribute to the Pu and Am concentratio

air? 

Why is code RATCHET, which was designed to reconstruct offsite doses from Hanford 

releases during its early years, used for onsite dose calculations at RFP, which is 

characterized by several closely constructed buildings? Wouldn’t these building wakes 

OCCU

1. 

ecord there is a notation of ppm Am. How was this value determined? How is 

2. 

ay

s for RFP? For what time period? 

prejudice the results?

PATIONAL INTERNAL DOSE (ORAUT-TKBS-0014-5) 

Page 84 of the internal dosimetry TBD shows a “Health Sciences Record.” On this 

particular r

it used in dose reconstruction?

It was noted on some of the records in Attachment C of the internal dosimetry TBD that 

some bioassay analysis was done offsite. What offsite vendors were involved in bioass

analysi
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is not known?

es?

5. 

6. nt of internal dose to unmonitored or inadequately 

monitored individuals?

7. 

. Which of the DOE complex-wide technical information bulletins were used in dose 

9. oluble

ss intake and dose for systemic organs when using bioassay and/or 

air concentration data? 

10. 

h

assess intake and dose for systemic organs when using bioassay and/or 

air concentration data? 

11. 

nvey

corresponds to X.X μm 

AMAD? 

12. d

3. How do dose reconstructors identify uptakes of Super Type S and other insoluble forms 

of plutonium, particularly prior to lung counting beginning in 1964? How is the dose 

reconstruction process modified for these cases? What default solubility values are 

applied, and on what basis, when solubility

4. Was any potential exposure from radon and/or thoron? If so, under what circumstanc

Section 5.3.1.6, Tritium, did not mention the form of tritium assumed in the dose 

assessment calculations. What is the default assumption for the form of tritium? Were 

there other forms of tritium present at the site? 

What method was used for assignme

There is no direction for dose reconstruction associated with neptunium, thorium, curium, 233U and 238

Pu. Has the potential for intakes from these radionuclides been considered?

8

reconstruction for Rocky Flats claims? Under what conditions were the technical 

information bulletins used? 

The TBD recommends that the dose reconstructor assume that radionuclides are s

for the purpose of dose assessment of the systemic organs and tissues. Is this approach 

recommended to asse

The TBD states that particle size and distributions are not available and it recommends

the use of the default value of 5 μm AMAD. On what basis is this default value 

recommended when a range of airborne particle sizes has been confirmed at RFP (i.e., 

don’t measured values take precedence over default values?)? Is this approac

recommended to

On page 9 of the TBD in Section 5.2.2.2, it is recommended that the “Dose reconstruction 

should assume a 5 μm AMAD, except for fires [for which] a X.X μm AMAD should be 

assumed for consistency with Section 5.2.1.3 above.” What did NIOSH intend to co

in Section 5.2.1.3 which does not exist in the TBD? What value

In general the values for median MDAs can be considered high, especially for enriche

uranium presented in Table 5.3.1.3.2, on page 14 of the TBD. What guidance is given to 

the dose reconstructor to deal with results below the MDA when assessing the missed 

dose?
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rial being cleared from the 

upper respiratory system, or from ingested material. The skin contamination contribution 

15. 

6. The TBD (Section 5.3.2.2.1, pp. 20 and 21) mentions that, “The method to detect 

the

at

is this

accomplished? 

18. o

e

es

? 

19. Were lung counting measurements a part of the routine monitoring program or was lung 

counting performed just in case of an accidental intake? On page 27, it mentions that 

“The dose reconstructor should note of the intake date. If the intake date is different form

the date for the Count Started, the intake date is from the file for a worker with a 

confirmed deposition. Otherwise the date of the lung count is used as the intake data.” If 

it was part of the routine monitoring program, why is the date of the lung count used as 

the intake date, which does not seem to be the most claimant-favorable approach? 

13. In Section 5.3.1.5.1 on page 16, the TBD states that until 1973, gross alpha measurements

were applied for analysis of the urine samples of workers potentially exposed to both 

uranium and plutonium. How dose the dose reconstructor evaluate potential dose 

contribution from neptunium and curium? 

14. In Section 5.3.2.1.1 on page 18, the TBD discusses two sources of interference. The 

second interference involves the contribution of the count from 241Am not in the lungs but 

from contributions from contamination on the skin, from mate

may be true for accident or incident situation, but not for the other examples, since the 

transit time in the upper compartments of the human alimentary tract is fast. How does

NIOSH deal with this contribution?

At the end of Section 5.3.2.1.2 on page 20 of the TBD, it states that “An underlying 

assumption is that americium remains associated with the plutonium particles in the lungs

until the particles are dissolved or removed from the lungs.” What is the basis for this

assumption? 

1

depleted uranium was to detect the 63-keV gamma (doublet) photon of 234

Th and to 

calculate the activity of 238

U, assuming equilibrium,” which was implemented manually

for special cases in approximately 1978. How did Rocky Flats Plants dosimetry 

personnel infer the lung activity of 238

U, due to depleted uranium exposure, based on

234Th measurement? 

17. In Section 5.3.3.1 on page 22, the TBD states that “Wound count information is largely 

irrelevant to dose reconstruction.” From other wound decontamination experiences it has 

been demonstrated that for some radionuclides, even with decontamination treatment, th

some wound contamination remains in the scars. In case of skin cancer, does NIOSH use 

wound contamination measurements for dose reconstruction and how

Figure 5C-16 on page 87, dealing with body counter information, shows that there is n

data for the right chest. On page 27, it states that “The dose reconstructor should estimat

the contribution for the right chest before using the data from the count because the lung 

data set generally includes contributions from both right and left lungs.” How do

NIOSH calculate the total activity in the lung in this situation
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be determined. What is 

NIOSH doing in its revisions to the TBD to expand upon this section? How is NIOSH 

ith the assessment of dose based on co-worker data? Will evaluation of

air concentration data be a part of this effort?

21. 

etion. Please explain the two different approaches used for uranium and for 

plutonium and how they were normalized in each situation? 

22. 

batch of samples.” On page 52, it states that “Not until 1973 were some

plutonium samples spiked with an internal tracer (first 236

Pu and, later, 242Pu). All 

plut the

median value of the MDA or the extreme value depends on the purpose.” This seems to 

ggest that some samples, in the earlier years, may have higher MDA than the ones 

tabu

dete ch spikes. What is the variability of the recovery? 

23. In A ed that

the alyte “Depleted uranium” and also for the 

alyte “Enriched uranium”. Is it possible for a worker to have been exposed to both 

24. ent A, on page 42 under the section entitled “Gross Alpha (1952–1971)” the 

D states that “The default condition, through 1963, was to credit the result to enriched 

e approach?

ed”

ments. What is the status of obtaining this key information and how are 

unmonitored internal intakes being modeled in the interim?

20. On page 30, it mentions that Section 5.5.2 is reserved and yet to

planning to deal w

In Attachment A, it is not clear how the volume of urine samples was normalized for 

24-hours excr

In Attachment A on page 39, it states that “Depending on the process, spiked samples, 

samples to which a known activity of the analyte was added, were generally processed 

with each

onium samples were spiked with an internal tracer after 1978...Whether to use

su

lated or described with the results of the urinalyses, since the recoveries were 

rmined by bat

ttachment A on page 37 in the Analyte/Method Code table on page 37, it is not

Method Code B1 appears both for the an

an

depleted and enriched uranium? If so, how does the dose reconstructor distinguish 

whether it’s DU or EU, once the method code (B1) has been assigned in both situations? 

In Attachm

TB

uranium unless the PHA count indicated otherwise...After 1963 (and enriched uranium 

operations were phased out), the default condition was to credit the result to plutonium.”

Why does NIOSH feel this is the most claimant-favorabl

25. In Section 5.5, Internal Dosimetry for Unmonitored Workers, it is indicated that for 

unmonitored workers “not involved in an incident,” use of air concentration limits in 

effect would be a reasonable approach. However, this section is left “to be determin

given that actual limits applied in the plant are needed, not merely those cited in official 

require
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CCUPATIONAL EXTERNAL DOSE (ORAUT-TKBS-0014-6) 

Dos

O

imetry Records

1. Page 10 states that further research is necessary to determine which dosimetry method 

of

2. the TBD, page 11 (and also ORAU-OTIB-0027, pp. 5–7), 

regarding skin and penetrating radiation doses, match that given in some of the historical 

Skin dose = hard gamma dose + soft gamma dose + x-ray dose + beta dose + 

+ x-ray dose) x 2.5 or 5.0 

depending on area and chemical form. 

ine skin and penetrating doses, but does not relate them

directly to the quantities as listed above which were in place at the time the doses were

3. d

4. 

he

early period (around 1951–1975) the deep gamma dose and the neutron dose were 

5. Was the Harshaw TLD 600/700 system used for extremity dosimetry during 1983–1991 

gamma and neutron dosimetry was being performed using the Panasonic 

system?

6. 

ort (NDRP) (Falk et al.

2005) assisted dose reconstruction (DR) during this period?

was used to record a worker’s dose. Has any progress be made in addressing this area

uncertainty?

Why doesn’t information in

records for the RFP (Owen 1964 and Putzier 1982)? In 1964, Owens defines the RFP 

doses as: 

neutron dose 

Penetrating dose = hard gamma dose + soft gamma dose + 1/3(x-ray dose) + 

neutron dose 

Hand dose = (hard gamma dose + soft gamma dose

However, ORAU-OTIB-0027, pages 5–7 uses the open window (OW), cadmium filter 

(CD), and brass filter (BR) to def

recorded. 

 

How is it intended that the dose reconstructor correlate the methods in TKBS-0011-6 an

OTIB-0027 in relationship to the recording methods that were in place during the years 

they were recorded, and also changed during the years?

How is it intended that the dose reconstructor reconstruct the separate doses to arrive at 

the total dose in these cases? Note on page 11, Section 6.3.1 that it states that during t

recorded as the total penetrating dose and not recorded as separate doses in the worker’s

file. 

when the beta/

Has any more information been found regarding the characteristics, performance and 

calibration of the LANL processed neutron track “plates” for the RFP from 1951 to 

1956? Has the recent Neutron Dose Reconstruction Project Rep
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7. e

4, some neutron dose 

was recorded, but the gamma and total dose was zero?

8. How are the dose reconstructors handling the possible manual correction needed and 

n,

mponents were not provided in the letter and, 

therefore, were not made?

9. 

10. Why aren’t uranium and non radiation workers included on page 18, Section 6.3.4.2, 

A film

e 

11. to SC&A the DOE-funded Job Exposure Matrix study by 

University of Colorado-HSC mentioned on page 18, Section 6.3.4.3, that appears to be 

12.

etimes worked in other facilities (beside the one their badge was 

calibrated to) on temporary or overtime assignments, which the dosimetry department 

nderestimate of dose to workers because of the change in the 

dosimetry response? Does this require some modification to the NIOSH decision stated 

for photon energy are needed? 

e of record rather than reporting this as a zero dose? Isn’t this inconsistent 

with the other AEC site TBDs and not claimant favorable?

Will the recorded neutron dose be used in dose reconstruction that is discussed on pag

14, Section 6.3.2.2.1 where it states that from July 1984 to Oct 198

discussed on page 14, Section 6.3.2.2.2, where it describes a possible manual correctio

but ends with the statement that dose co

Why isn’t more information provided besides the exchange history (i.e., quarterly dose

history, etc.) that is discussed on page 18, third line list: “1977 – present, dosimeter

exchange history?” 

where it states that the NDRP (Falk et al. 2005) will cover 1951–1970 when NT

was used for neutron dosimetry? How is NIOSH handling dose reconstruction for 

workers exposed to other neutron sources [(alpha, neutron) reactions such as in UF4, 

criticality experiments, calibration sources, etc.] besides PU at the RFP during this tim

period when NTA film results/procedures are in doubt?

Can NIOSH make available

helpful in dose reconstruction? 

How has NIOSH dealt with the situation discussed on page 19, Section 6.4.3 that 

individuals som

could not detect? Does NIOSH consider this an occurrence that happened often enough

to create a significant u

on page 27, Section 6.6.3.4 that assumes that each worker stayed in the work area where 

the badge was calibrated and therefore no corrections

13. Why isn’t the 10 mrem minimum detectable dose discussed on page 20, Section 6.4.4, 

used as the dos

Common Issues

14.

not be a date associated 

with a dose record? Also, page 15, Section 6.3.3.1.3, mentions that the Activity Date 

Page 21, Section 6.5.2, second paragraph states that “If no activity date is associated with

a dose record...” What is an “activity date”? Why would there

may be outside the employment period. What does this mean?
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15.

Photon Dose

In the dose reconstruction process discussed on page 23 regarding exposure geometry, 

why does NIOSH consider the assignment of ISO or ROT instead of AP geometry as 

always most claimant favorable (i.e., for anterior-seated cancers)? 

17. dose reconstruction the 

effect of shielding where photons with energies >250 keV may contribute more to the 

e dose because of their greater penetrating power? The TBD states on 

page 24, Section 6.6.1.1, that the photon exposure spectra, assumed to be mostly in the 

e. 

 

18.

etry response increases with angle, but on 

page 28, it states that the Panasonic system response decreased as the angle increased. 

19. Why is the uncertainty for low doses of 1 and 2 mrem less than for the higher doses of 

t is

21. e

22.

16. On page 24, 6th line up from the bottom, shouldn’t it read ...the 30-250 keV photon..., 

instead of ...the 30-50 keV photon....?

Has NIOSH considered the need to take into account during

final effectiv

30–250 keV range, is claimant favorable because of the higher REF for this energy rang

Could NIOSH provide more information on how angular dependence has been 

addressed? On page 27 it states that the dosim

The Y-12 ORAUT-TKBS-0014-6, page 18, states that the recorded dose of record is 

likely too low at non A-P angles for beta/photon doses. What is NIOSH’s position on 

angular dependence for: (1) NTA film, (2) beta/photon film, (3) TLD neutron, and (4) 

TLD beta/gamma badges?

photons (i.e., 1 x1.00 = 1 and 2 x 1.00 = 2, but 10 x 1.23 = 12 and 100 x 1.23 = 123 

mrem, etc.) in Table 6-12, page 30? Tables 6-10 and 6-11, page 29, show the opposite

characteristics. 

20. Please explain how Table 6-3 will be used by the dose reconstructor to amplify wha

provided on page 21?

In calculating the exposure geometry factors, what dose rate and exposure durations wer

assumed? From what source was this data derived?

Did RFP ever acquire any sealed sources for use in industrial radiography such as 226

Ra, 

137Cs, or 90Sr?

Neutron Dose

23. Why is there a difference in the limit of detection (LOD) between page 33, Table 6-16, 

which lists the limit of detection (LOD) for neutron NTA film for 1961 as 96 mrem and 

Table 6-17 on page 34, where the LOD is listed as 120 mrem?

24. Page 34, Table 6-18 list the potential missed neutron dose below 800 keV for early NTA 

film to range from 16% to 60%. However, the text below it selects 56% as the claimant-
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hy wasn’t the higher value

of 60% missed neutron dose with a resulting multiplying factor of 2.5 selected (i.e., 1/1-

0.60) = t of

neutron 2.27 instead

of 1.79

25. Page 35 eport

is finish ry 7, 2005. Is

the ND ificant

changes in the results of past or current dose calculations? 

Electron D

favorable value with a resulting multiplying factor of 1.79. W

x 2.50)? In addition, even if the value of 56% was selected for the percen

dose missed, why wouldn’t the multiplying factor be: 1/(1-0.56) = x

?

, Section 6.7.3.4 states that other information will be used until the NDRP r

ed. The NDRP report (Falk et al. 2005) was available as of Februa

RP currently being used in dose reconstruction? Has it resulted in any sign

ose

, next to last sentence states that the RFP had problems with elevate

om contamination on leather gloves worn during foundry operations. Di

26. Page 37 d beta dose

rates fr d these

orkers routinely wear wrist badges so that these doses could be accounted for? If not, 

27. Pag

bad

mig

w

how will dose reconstruction be performed? 

e 40, Section 6.8.4.1, addresses angular dependence and states that the assembled 

ge displays severe angular dependence to beta exposure and that for TLD badges it 

ht record only 36% to 59% of the true beta dose at +30 degrees. Wouldn’t this need 

e addressed, especially in the extremity section where the worker’s normal 

vements would not tend to average it out? 

e 41, Table 6-24 footnotes states to multiply any dose greater than 2 mrem by 1.12.

ever, the forth column contain numbers that do not use this multiplier and do not 

to b

mo

28. Pag

How

match Columns 2 and 3. Why is this?

29.

nly

d isotope, 238U at 3.40E-07 Ci/gm of 

DU is not itself an active beta emitter)]. Page 37 states “Thus, for depleted uranium, one 

with 2.29-MeV (Emax) beta particle from 234mPa, the most energetic 

contributor to the beta exposure.” 

Un

How can VARSKIN be used if it leaves out 234mPa? Page 42, the last paragraph of 

Section 6.8.6 recommends using the VARSKIN software to calculate skin dose from

contamination. However, Table 6-26 on page 43 shows that the VARSKIN Mod 2 o

includes one (234Th at 3.40E-07 Ci/gm of DU) of the two most active beta isotopes [the

other being 234mPa at 3.40E-07 Ci/gm of DU (the thir

is dealing essentially

monitored Individuals

30. at

sho

195

mo

What other measures, such as the NDRP, are being implemented to identify workers th

uld have been monitored, but were not? Page 42, Section 6.9.1, states that in the early

0s the only groups expected to receive doses greater than 10% of the RPG were 

nitored.
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rs included most of the employees 

working in Building 71 (now Building 771). Only a small number (10–18)

neutron reaction with light elements, especially from the plutonium 

b. 

953–1984. Another RFP document entitled Film Procedure Timeline for the 

Neutron Dose Reconstruction Project 2003 (RFP 2003, pg. 6) provides the number of 

of the total workers badged, on the average only approximately 25% 

were issued NTA film badges (over the 10-year period, it ranged from 9% to 50%). 

proceeded per month (therefore, 300 workers 

were badged for neutrons throughout that year), this results in a ratio of 300/1362 = 

cts raise the question as to whether the policy and procedures in place during this 

period at the RFP was adequate to ensure that the workers who were at risk of receiving

significant radiation doses (especially neutron doses) were actually, and adequately, 

badged. It would appear that the “10% of the RPG” policy did not necessarily provide 

adequate monitoring for all workers that needed it. 

Extremity Dosimetry

a. The NDRP report (Falk et al. 2005), page 1, states that: 

For some plutonium workers, neutron monitoring was not provided until 

the early 1960s, and their dose of record may not include significant 

contributions from neutron exposure received prior to being issued a 

neutron dosimeter. These worke

of these employees were monitored for neutron exposure, and that 

monitoring was only during the period October 1956 to September 1957. 

Operations in Building 71 involved chemical processing of plutonium in 

acid solutions and resulted in significant neutron fields from the alphatetrafluoride compound. No evidence has been found that neutron 

shielding was present for these operations until the early to mid 1960s. 

Column 2 of Table 2, pages 17 and 18, of the RFP annual radiation exposure report 

for the year 1984 (RFP 1985) provides the total badged personnel per year for the 

years 1

NTA films processed per month. Analyzing the two tables for the years 1959–1969 

show that out

For example, in 1960, a total of 1362 workers were badged throughout the year and 

approximately 300 NTA films were

22%. 

 

These fa

31. Is NIOSH currently working on TBD changes to make more clear how the following 

issues are addressed in, Section 6.10, page 43? 

a. NTA film in wrist badge - Between 1951 and 1970 film dosimetry was used for 

extremity dosimetry. Was this only for beta/photon? Was any NTA film used in the 

wrist dosimeters? If not, how will the extremity dose from neutrons (which could 

have been greater than the whole-body dose for those workers handing neutron 

emitting material) be reconstructed?
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it

be considere y (WB) dose to the

hand if he/she was working on a task re [The NDRP (Fax 

nly 40% of the wrist

the details on the hand-toow can 

alues (none were presented in the TBD) or how they were derived?

 - The last sentence in this section states that additional 

ormation on the hand-to-wrist ratios is required before shallow doses to the 

ities can be reconstructed. The last paragraph on page 37 states that the beta 

work

p . Has any additional information been acquired and is it being 

SUPPLEMENTARY EXTERNAL DOSE INFORMATION FOR ROCKY FLATS PLANT 

ird line: shouldn’t this read Rocky Flats Plant – Occupational 

metry (Rev 01).

2

? 

dele

b. Hand-to-wrist-to-WB ratios – If a worker did not wear a wrist dosimeter, how can

d claimant favorable to assign only the whole-bod

quiring hands-on work?

2005), page 22, list the average WB photon dose response as o

photon dose. Additionally, in most cases the dose to the hands would be greater that 

the dose to the wrist (Mann 1964).]

c. Valid hand-to-wrist ratios - It states in this section that

wrist ratios are not available. In a letter by J.R. Mann (Mann 1964) it is stated that 

the hand dose is 2.5x or 5x the wrist dose, depending on the work location]. H

it be determined if these ratios represent the working environment at RFP if we do 

not know their v

d. Addition information

inf

extrem

dose to the hands could have been a problem because of considerable hand-on

erformed at the RFP

used?

(ORAUT-OTIB-0027) 

1. Page 5, first paragraph, th

Environmental Dosi

Page 5, in the pre-1960 section list that skin dose was determined by (OW + CD + BR)

and also by (OW + CD). Isn’t the BR term in error in this paragraph and shouldn’t it be

ted?
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ATTA

Dates: Sep

Sep nal Medical Dose TBD 

Sep

R

t

NIOSH , Sam Glover 

Additional

Introducto

General C

SC&A: s since we started our review of the RFP TBDs, we have boiled down 

SI

SC&A: 

ere

CHMENT 4: SUMMARY OF CONFERENCE CALL ON SC&A 

QUESTIONS PROVIDED TO NIOSH 

tember 6, 2005, General Questions and Site Description TBD 

tember 8, 2005, Occupatio

September 6, 2005, Occupational Environmental Dose TBD 

September 7, 2005, Occupational Internal TBD 

tember 7, 2005, Occupational External Dosimetry TBD 

FP General Questions/Site Description TBD Conference Call Summary

Date: September 6, 2005 

Time: 10:00 am - 12:30 am EDT 

Participan s:

ORAU: Bob Meyer, Craig Little, Jim Langsted, Roger Falk, Theresa Lopez, Laura 

McDowell-Boyer 

: Brant Ulsh, Greg Macievic

SC&A: John Mauro, Joe Fitzgerald, and Tom Bell 

Participants: Amy Dean (Scribe), Karin Jessen, Al Wolff 

Advisory Board – Mike Gibson 

ry Comments: 

NIOSH: We are hoping that we can go through the SC&A questions methodically and deal 

with them in detail. Some we will be able to answer, some we can perhaps resolve 

now, and some will need to be resolved later. 

omments: 

In the 3 week

our questions into talking points in a PowerPoint presentation that might help to 

clarify and elaborate on the questions. We hope everyone has a copy of these. They

will form a good basis to review our overall preliminary issues. Joe Fitzgerald has 

received NIOSH’s input on their initial comment resolution which has been placed

within a table matrix. This matrix has been very helpful. Realizing there was a 

holiday during this period, we appreciate you getting this data to us so expeditiously.

DISCUS ON OF THE POWERPOINT SLIDES 

Slide 2, Bullet (2). The excess Pu and HEU storage vulnerabilities are not addressed 

in the TBD. However, this was addressed by DOE in the 1990s. Assessments w

done in Pu vulnerability and they did assessments as well on where the Pu was being 
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ial for

ORAU: and

d right now on the Argonne report, but we do plan to go back and address this. 

We have addressed this from the cleanup program perspective as best as possible; 

SC&A: Our interest is in the potential exposures. 

NIOSH:

SC&A: Will the assessments and reviews that were done in the 1990s be reflected eventually

NIOSH:

rmation in rev

summary of what you can provide us. 

SC&A: In the evaluation of issues in the SEC, they have asked: “Why are these 1990s 

aks, there is 

not much new that is important. But we realize that this needs to be extended and 

ey existed, we did not have the time to go through them

before this. 

ORAU: In the 1980s, nothing of a large scope occurred. In claimant files, there is 

excruciatingly detailed information about each incident in which they were involved. 

agnitude of 

e of the customary particle size. 

uction.

ressed.

y of the

uld

roven otherwise. 

stored. RFP held 96 MT in form of pits and spares which provides the potent

criticality issues. Also, accumulation of combustibles and potentials for fire were 

abundant. During the post-1989 cleanup, there was a lot of potential for exposure to

workers involved in waste management. What are NIOSH’s plans to characterize 

these exposure histories?

We did not address this in the initial go around but now we are trying to go back

see if we can understand if there are any major events that will affect dose. We are 

burie

including the D&D effort and this includes what badging was done. 

This will be considered in the TBD Site Description. We realize that the TBD does 

not characterize this yet.

in the next round of TBDs?

If you have copies of these assessments and reviews, please provide anything you 

have that will help us. We will be adding additional info isions to the 

TBD and would be glad to include a 

reviews not used to the period from 1987 to 1989?” Also we note the occurrence 

reports were missing. 

NIOSH: Some incidents are actually covered in Tasks 5 and 6. After our review of the 1969

fires and what was left over from the Building 903 drum storage area le

anything important addressed. We have found 27,000 pages of records that are a new

find. Although we knew th

The claimant’s file has an overabundance of data that is helpful in doing dose 

reconstruction. We did address in the TBD the 1957 Pu fire and the m

their intakes. The only data we have indicate the us

SC&A: It sounds like there is a confluence of new data that is relevant to dose reconstr

With RU you have a potential incident where Np has not been adequately add

There appears to be a wealth of new incident data which now enriches the stor

transients in exposure. Some additional radionuclides may be around and co

contribute to dose even though their contribution may be low until p
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NIOSH:

ata.

is just a supplement to what ORAU already has. We do realize 

that we need to look more at RU. Roger Falk is working on this and what he finds 

r

NIOSH: We realize that there are other issues to pursue. We need to get more data on how 

s information, we may want to

SC&A: up

z. This should include further discussion of RU, 233U issues, and how 236U 

NIOSH: 

SC&A: Perhaps we could have Bob Alvarez dial into the end of the last part of the internal 

NIOSH: That would be even better, as it will be hard to find another time to get these folks 

1. W

ra

ORAU: Section 6.9.1 explains that in the early 1950s only groups expected to receive 

rsal prior

to 1964. In that year, the security badge was incorporated in the dosimetry 

badge, which ensured that each individual wore a dosimetry badge 

(Putzier 1982). This design was maintained until the early 1990s, when the 

Missed dose for worker intakes is based on what is in the worker’s file. The 

additional files will not affect this data for the individual claimant who has this d

The additional data

will be captured in the next revision to the TBD. But it is important to point out that

only a small part (21 MT) of the total 2373 MT is applicable. Therefore the 25% of

Pu that gets into the bone or liver may not be significant. But we realize that an 

argument should be provided in the TBD revision as to whether this is important o

not. 

SC&A: Could your current methodologies miss some things that the new records might help 

to clarify?

IAEA handles co-worker data. If we find some of thi

put that in the TIBs. RU will be pursued. Characterization of monitored vs. 

unmonitored will also be pursued. 

We need to find a time to get Bob Alvarez on a phone call with you in a small gro

at a later time to discuss this in more detail. This would be very helpful to Bob 

Alvare

figures into RU operations. 

We will send out an e-mail on Thursday or Friday to make this happen. 

dose conference call tomorrow. 

together again, and they will all be on the internal dose call tomorrow. 

SC&A: We will call Bob Alvarez and try to set this up. I think he may be planning on being

a part of that call already.

ORAU General Questions: 

SC&A:

hat were the criteria for monitoring workers for external penetrating 

diation? How did these criteria evolve over time?

 

doses greater than 10% of the Radiation Protection Guideline would receive 

dosimeters. As explained in Section 6.4.1, dosimetry was not unive
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be

got

ORAU: It is difficult to go back that far. That level of detail is hard to come by. We have had 

For a dose reconstruction, how does ORAU do this?

NIOSH: ose,

individual’s dose. We look at

the person’s job title. If they worked in non-radiation areas, they were given the 

orker

SC&A: We realize the University of Colorado Job Exposure Matrix might be very helpful in 

NIOSH: We have been working hard to get the University of Colorado Job Exposure Matrix 

SC&A: re Matrix?

SC&A: So are you just in the minimum/maximum mode now on addressing RFP claimants?

NIOSH: If the claimant’s dose puts him over 50% probability of causation (POC), then the 

security badge was separated from the dosimeter and individuals unlikely to 

receive occupational radiation exposure greater than 100 mrem were no longer 

issued dosimeters.

SC&A: The TBD is a very good product, but we do have some questions. There seem to

important juncture points where policy changed and dosimetry types changed. We

would like to better understand how it was decided who the workers were that had the

potential to exceed 10% of the guidelines. Who made this decision as to who

badged and who did not?

to rely on a few pieces of documentation. We don’t understand how this 

documentation was done. 

SC&A 

We take a claimant’s case. If they have dosimetry, it is easier. If they have no d

then we have to use co-worker data to determine the

environmental dose. If they are radiation workers, then we have to rely on co-w

data. 

this regard. What is the status on NIOSH getting access to the Job Exposure Matrix? 

We realize there is a problem on who actually owns that data. 

from Dr. Ruttenber and our lawyers are working on it. 

Are you hampered without it being available to use the Job Exposu

NIOSH: Once we get it, we will use it. We have been getting along without it, but when we

get it, we believe it will help. 

minimum/maximum mode makes these easy compensation cases. If they are 

administrative worker cases, it is likewise easier to determine the potential dose and 

the POC. When they get to claimant cases where the POC is near the 50%, we are 

going to make some hefty overestimates. 

SC&A: You pointed out in the TBD that those who got badged and those who did not and 

still worked around radiation in the early 1950s is often uncertain. 
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r data.

re

ssed dose, do you use the 

geometric mean and the geometric SD? Or in other cases, do you use the 95th

s not always clear which is used.

SC&A: 

2. What percentage of Rocky Flats Plant (RFP) workers were measured for external

ra

ORAU: The data necessary to develop these statistics have not been located. Based on 

SC&A: We are trying to get a feel over time, what percentage of workers was monitored over 

rnal

dosimetry this morning, it is not clear. 

NIOSH:

ate. We do have for external dose, the years they worked and percent 

of workers monitored. But for neutrons, this is not as easy to get. 

SC&A: to have the information on the claimant population, but this may not be 

true for the whole work force. What percentage of workers was monitored in the 

ORAU: red.

rker

NIOSH: but we will get them. We think the more important 

question would be: “Can it be determined if the worker was likely to exceed the 

claimant has a likely exposure. But if we don’t know who might have exceeded 10% 

and this was missed during the dose reconstruction, that would not be conservative. 

NIOSH: There is a TIB on how to use co-worker data. It is OTIB-0020 provides guidelines on 

how to use co-worker data to determine dose to unmonitored workers. This OTIB 

provides a lot of information on how to utilize co-worke

SC&A: Does OTIB-0020 deal with surrogate data? Do you use the real distribution for 

people who were not badged and need an environmental dose? Or do you use the 95th

percentile rather than the natural distribution for high dose potentials? When you a

using your co-worker data as a surrogate for this mi

percentile? When we look at the TBD, it i

NIOSH: You would use the known data if you have it. If you do not know the job and they 

are radiation workers, then we might go to the 95th percentile.

diation over time, including neutrons and photons?

the previous comment response, it is likely that the overwhelming majority of 

workers were monitored during the lifetime of the facility. 

time for neutron and photons. In the NIOSH matrix data that we got on exte

The matrix that was completed on Friday, September 2nd, and provided to you is 

already out of d

You seem

1950s and 1960s?

Workers who were likely to receive 10% of the guidelines were monito

SC&A: Based on the 10% rule for badging, how many workers were badged in the mid to late

1950s? What percentage of the workers was monitored? When you look at co-wo

data, is it based on a relatively small population? If it is only 5% or 20% of the work

force that is different than if it was 60% of the work force. 

We don’t have those numbers,

10%”. If we can, then this is good. We could then use other co-worker data if that 
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ing, it is likely that the percentage of those workers who were badged were 

low and that this percentage likely increased with time. 

ORAU: 

worker groups that included the individual. For each of the operations, workers were 

SC&A: 

ORAU: 

rted to combine the security and FB in 1964, at 

least for the photon doses. When the TLD came in during 1974, it became even more 

SC&A: In work groups, such at Pu workers, was there this same evolution as with the 

individual workers mentioned above? 

ORA

NIOSH: We will include whatever data we can develop on this and get it to SC&A. 

ORAU: d.

SC&A: 

ternal dosimetry dose reconstruction is OTIBdeveloped as well just for RFP that will be 

ORAU: 

he gap for missed dose will likely be more for administrative people 

ing a

NIOSH: co-worker data even if there is only a possibility of exposure. 

SC&A: Is there a way to look at the percentage of workers who were badged? When the RFP

got runn

The 10% rule was not applied to the individual worker. It was usually applied to 

badged the same way. The rule would apply to the workers in buildings and areas 

where they are likely to exceed the 10% guideline.

But, were they applied very consistently?

This evolved over time. In 1958, for Pu workers, badging was done according the

10% rule. It as probably also applied to the EU workers. But this would have been

phased out in the mid-1960s. We sta

consistent. 

U: For U workers it was consistent from the very beginning. For Pu workers it is not as

clear. This is very relevant when you are doing dose reconstruction. Those 

perspectives are in the TBD, but the distributions over time are important. 

We can’t determine the fraction of the work force that is not occupationally expose

It is not likely that we can get this kind of data for any period of time, especially for 

the early time. 

When you do the co-worker analysis, that kind of information over time would be

helpful. 

NIOSH: The general use OTIB for co-worker ex

0020. There will likely be a new OTIB

even more helpful in doing occupational external dose reconstruction. 

But, you are missing the point. The unmonitored worker who is not badged is going

to be rare. T

where it is not known or suspected that they did enter an area of potential exposure. 

But for unmonitored radiation workers who you know are likely to have been in a 

potential area of exposure, you can rely on the co-worker data when you are do

dose reconstruction for them. 

We might need to use
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at

NIOSH: Well, we don’t have access to the Job Exposure Matrix yet. If workers are not part of 

ker data might still be helpful 

d

NIOSH: Once we get the Job Exposure Matrix, we will assume they are in the group and this 

How are unmonitored workers addressed, particularly before the 1964 transition to a 

combined security/dosimetry badge? For example, guards were apparently not 

ction areas;

RAU: Dose Reconstruction procedures to address unmonitored workers are applied to this 

ot routinely monitored. The dosimetry 

data file provided by DOE for each claimant does indicate the unmonitored periods. 

are not a part

the 10% group, but after that it is

NIOSH: They would use ambient dose from the external TBD, but it could also be handled on 

a case by case basis. 

SC&A: If the worker is working in vaults with the pits, this could result in a hefty dose. 

SC&A: So there will be a new OTIB, like one done for Y-12 and the analysis will be in th

OTIB?

the radiation worker workforce, the co-wor

SC&A: The general approach is that groups of workers were either in the program or not in 

the program. If they were in the program, they will have some information and if

they are in the 10% group, then they are followed more closely. As time goes on an

more and more groups were turned on, it is likely that more workers and groups will 

be included in the rules for the 10% population. Of course, once the 10% rule was

turned on, then everyone would have dose data. If you were missing a dose for a 

person, then the group dose would be used as the surrogate. But the job exposure 

matrix might create an incongruity. 

will then allow a conservative estimate. 

SC&A: 

3. 

routinely monitored, although they were often stationed adjacent to produ

how would their exposures be characterized?

O

situation. It is not clear that guards were n

Dose assignment is based on job description, era, dosimeter history, and interview 

information. Guards that were likely outside of production areas are often assigned 

ambient per section 6.9.2 of the External dose TBD. Each case is considered 

separately. 

SC&A: Were the guards routinely monitored?

ORAU: No specific information on this has been found, but it is likely that they

of the 10% group. In 1964, they were included in

not certain. 

SC&A: You would then have to assign a co-worker dose from another group. Has this been 

made clear in the guidelines for Task 5? Who handles unmonitored workers?
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We would treat guards as radiation workers, if there is enough information. The 

guards would have had access to go into various areas of the site. If it is likely their 

tion workers.

IOSH: No, it more often is a summary data for the whole building or site. That data shows 

at

e. The individual judgments of the dose reconstruction as to 

whether or not the claimant is working with badged workers or not is important too. 

When doing a minimum/maximum determination, we would apply the 95th percentile. 

When applying the OTIB, it is not being used for best estimate dose. The 

ht use the geometric mean?

SC&A:

4. What data did NIOSH/ ORAU obtain regarding the degree and extent of 

co

p

ORAU: 

C&A: Bob Alvarez was interested in this subject, but could not be on the call at this time. 

ORAU: 

SC&A: 

ORAU: 

This is a good document to pull out and review. 

C&A: Maybe we can go over this later. This also applies to the 232

U component as well. 

ORAU: ent to this in our responses to questions 5, 6, 7 and 8 below. 

NIOSH:

jobs would take them into radiation area, they would be treated as radia

SC&A: Is co-worker data available by job group?

N

the number of badged or unbadged and how it is distributed by different locations. 

This kind of data does not show it by individual job title. We apply the geometric 

mean and the geometric SD and apply a 3.0 uncertainty. But it does depend on wh

kind of data they hav

minimum/maximum is just used for overestimates. 

SC&A: But you mig

NIOSH: It depends. 

ntamination of specific processes that resulted from special projects such as 

rocessing/fabricating of 233U and other radioisotopes?

Roger Falk will have to address this later. 

S

There is a Word document that was sent to Bob Meyers that gives Roger Falk’s

comments. Joe Fitzgerald should have that response. 

We did not see that particular reference. 

It should be on the O Drive. If you don’t have it, we could provide it. 

NIOSH:

S

There is something pertin
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SC&A: 

5. How much U was handled at Rocky Flats over what periods of time?

ORAU: 

C&A:

6. What were the levels of U contaminants present in the U processed and handled at 

R

emove the 

Th-228 and its daughters, which were troublesome mainly because of their 

contribution to elevated external doses. 

al

32 contaminant with Th-228 and its daughters 

ingrown at 90 days after the thorium strike. For type M solubility, assessments based 

s, except for 

eceived an

orate 

h we

all, we had really

ed like 

omething about its chemistry, was much easier to contain 

and clean up as compared to plutonium.” 

SC&A: 

Why doesn’t the TBDs account for recycled uranium of various assays that were 

processed at Rocky Flats?

ORAU: 

epartment of Energy's Rocky Flats Plant, 

1953–1993," June 30, 2000. It was determined that a very small fraction (0.03%) of

the depleted uranium processed at RFP was known to have resulted from recycled 

233

We have no comprehensive data regarding the amount of U-233 handled. TBD 

section 2 indicates that small quantities of U-233 were handled in Building 881. 

Approximately 20 kg of U-233 was handled as a special project at Rocky Flats in the

mid 1960s, according to Ed Putzier in his memoirs, “The Past 30 years at the Rocky 

Flats Plant,” pp. 28–29. This document is in the Site Research Database with the 

Reference ID 4632. 

S

232 233

ocky Flats?

ORAU: Per the reference cited in Comment 5, approximately 40 ppm U-232 was in the U-233

mixture. A thorium strike was performed after receipt of the material to r

We have performed calculations based on this data to determine the possible intern

dose impact of this level of U-2

on U-233 only will account for 80% or more of committed doses to organ

bone surfaces, bone marrow, gonads, and liver. 

However, there is no evidence yet found that any worker at Rocky Flats r

intake of this type of material. Ed Putzier cites, “...there were very elab

precautions made in order to campaign this material. The machine tools whic

used were heavily shrouded with plastic to protect them and, as I rec

no trouble in containing the alpha radioactivity. I am not sure why but it seem

the uranium, maybe it is s

7. 

Recycled uranium was carefully reviewed by DOE and documented in, "Report on 

the Flow of Recycled Uranium at the D
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ented Pu, Np, and Tc levels in 

inuscule fraction (0.0003) of DU 

for

contribution of the recycle contaminants to committed organ doses 

SC&A: s some interest here. 

 

SC&A: 

SC&A: 

8. W

co

The DOE Recycled Uranium report identified two processes that had the 

from Fernald indicated this is a potential 

of

These two processes with RU were cited in the TBD, but we would like to see more 

SC&A: 

9. cled uranium measured for exposure to 

contaminants such as 239

Pu, 237Np, and fission products? The TBD appears to have no 

in

uranium processing, and this material contained plutonium, neptunium and 

technetium below de minimis levels.

We have performed calculations based on the docum

recycled DU and HEU at Rocky Flats. For recycled DU and for both Type M and 

Type S solubilities, committed organ doses assessed based only on non-recycled DU 

accounted for over 93% of the dose for all organs except bone surfaces (75%), red

bone marrow (88%), and liver (79%). Given the m

at Rocky Flats of recycled origin and that many DU workers were also monitored

Pu exposures, this is not viewed as a significant internal dosimetry issue. For 

recycled HEU, the

is less than 0.00000001, again not a significant internal dosimetry issue. 

Then your bottom line is that there i

ORAU: NIOSH had us stop work on that OTIB. Liz Brackett could give us a status. 

We have a lot of interest in RU. Irradiated rocket fuel was passed through RFP. 236U 

came out of that. In the 2000 RU report we note that RFP was involved. 

hat processes at Rocky Flats would have been likely to concentrate contaminants

ntained in recycled uranium?

ORAU: 

potential for concentrating or releasing transuranics or fission products related 

to worker exposure. These were the vacuum melting and casting processes 

for DU. Although information

concentration point, more recent data from the Specific Manufacturing 

Capability Project indicates that no accumulation occurs. The conversion

DU to oxide in the RFP "chip roaster" was identified as a potential 

concentration point, but was not supported by associated emissions analysis 

for transuranic elements. 

SC&A: 

on how this is characterized for the dose reconstruction. 

Were workers who handled and processed recy

formation about this issue of concern. 
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ORAU: 

ossible or

actual work assignments in plutonium operations. 

ORAU: 

d Np, there is a dose contribution,

IOSH: We do need to point out what is significant if RU is part of the dose distribution. It 

ith this in the revised TBD. 

su

SC&A: 

11. H

R many were not monitored for internal and 

external exposure? Of those monitored, how many (on a fire-by-fire basis) 

received doses in excess of prescribed limits? 

ORA

x drain fire and does

indicate the number of workers unexpectedly exposed above one lung burden. 

It is unlikely that special dosimetry was utilized for workers that processed recycled 

uranium. Based on the low level of contaminates in this material and the lack of 

knowledge about this at the time, special dosimetry was probably not used. There is 

no documentation of monitoring for contaminants. 

No, uranium workers were not monitored for recycled U contaminants as such. Many

uranium workers were monitored for plutonium after 1960 because of p

NIOSH: For Pu workers, the missed dose would already have been applied by the Task 5 

group of people. 

SC&A: We feel you should look more into the trace elements. However, the dose 

contribution from fission and activation products that come along with RU are not 

likely to be important contributors. 

We did some of the scoping calculations. It depends on the organ. It makes a 25%

difference for the bone and 21% of the liver. Pu and Np are only partially 

characterized in the worker dosimetry. For Pu an

and for the bone it is important and can range up to 25%. For the others organs, it 

might only make a 2% difference. 

N

should be addressed in the TBD, as it was done in Hanford. We will incorporate 

Roger Falk’s calculations in the TBD and deal w

SC&A: 

10. What radiological controls initially were established at Rocky Flats and how did they 

bsequently evolve over time?

ORAU: We’re not sure of the intent of this question. Does the term radiological 

controls refer to physical barriers, such as gloveboxes and shielding, or 

monitoring processes? All data available at the time the TBD was written are 

summarized in the various chapters.

ow many workers were exposed unexpectedly to radioisotopes from fires at 

ocky Flats? Of this group, how

U: Section 2.6 of the RFP Technical Basis Document addresses accidents at RFP, 

including fires. Section 2.6.2 discusses the 1965 glovebo
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fires.

routine

ible

n Dosimetry annual 

reports produced in the mid-1980s (cited elsewhere in these questions) provides these 

have been

found. Roger Falk will see if any additional information can be found. 

SC&A: 

fficult to determine which of the worker groups were 

ORAU: for intakes. They were the same people who 

the intake they received by inhalation. All 

fire fighters were monitored for possible intake. 

SC&

at it is not available.

is

They did not work up a summary report for a whole group in these situations. They 

put this information in the individual’s files. 

ORA a

ent up the stack. It also occurred off shift 

blown

rs near

these fires. It is likely that the onsite air dose would 

include this in the air calculation. Onsite ambient air concentrations, with or without 

fires, should account for this. In 1969, the filters stayed in tact and the release, 

SC&A: 

d

Unfortunately, we know of no compilation of this type of information for other

Any workers identified as unexpectedly exposed would be subject to nonbioassay, which would be recorded in the individual's dosimetry file. The "Poss

Inhalation Cases Body-Counted" table included in the Radiatio

numbers for a limited number of years. No previous or subsequent reports

In terms of fires, there is a lot of interest in high-fired oxides. But with so many 

smaller fires, it will be di

monitored and which were not. 

All the people affected were monitored 

were monitored for external radiation exposure. This would cover the fire fighters in

the first fire in 1957 in Building 71. Internal dose for the support people is 

characterized, but for the external dose we are not so sure. This, however, is not 

much of an issue. The important thing is 

A: This kind of information is available for big fires. But for the smaller fires, we are 

concerned th

NIOSH: We think we will have to rely on the individual worker files where bioassay data

likely to be recorded for such events like fires. 

ORAU: You can also rely on some of the follow-up data as well. 

NIOSH:

U: What about the people who were walking by? Only in 1957 and 1969 would this be

problem. In Building 771, however, it all w

in the evening. Most of the release was in the first two hours, and would have

away at a height that would have precluded having much on an affect on worke

the building. Therefore this is not important. John Till’s report in late 1990s 

provides a lot of information on

therefore, was much smaller. 

12. How many workers entered airborne radioactivity areas without respirators and how 

id they assess or estimate this worker’s inhalation dose?
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Records summarizing the numbers of such incidences have not been found. If 

a worker was reported to have entered an airborne radioactivity area, the 

iggering special bioassay,

which would be reported in the individual's dosimetry record. 

SC&A: 

We can’t answer that. We have never taken into account the wearing of respirators. 

We just assume they did not use them and did our dose calculations as if they were 

SC&A: How many cases do you have where surrogate air concentration data was used rather 

NIOSH: entioned. It is possible that 

Mr. Robinson in Task 5 could answer this. We had asked him to be on line but he 

s.

SC&A: Let’s defer discussion on questions 13, 14 and 15 to a later time when Bob Alvarez 

14. W ve to drains and

overflows that may have evolved over time?

ORAU: 

SC&A: 

15. What parameter controls (i.e. mass, enrichment, density, concentration, 

g

a cky Flats and how did they evolve?

RFP. The nature and extent of the criticality safety program was not reviewed 

in the preparation of the TBD. 

ORAU: 

incident was handled as a "Potential Intake" event, tr

Are there procedural records that characterize the wearing of respirators?

NIOSH:

not wearing respirators. 

than bioassay co-worker data?

In applying OTIB-0020, we don’t remember if this is m

apparently was not able to make it. 

SC&A: 

13. How many workers required chelation therapy during the history of the operation of 

Rocky Flats?

ORAU: Information on specific workers would be contained in their dosimetry record

Roger Falk and Joe Furman will research this. 

can be on a call to discuss these three questions.

SC&A: 

hat criteria and guidance were used to verify criticality safety relati

There is no evidence that an accidental nuclear criticality has ever occurred at 

RFP. The nature and extent of the criticality safety program was not reviewed 

in the preparation of the TBD. 

eometry, interaction, reflection, moderation, and neutron absorption) were 

pplied to prevent criticalities at Ro

ORAU: There is no evidence that an accidental nuclear criticality has ever occurred at 
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16. D

ORAU: 

SC&A: 

ORAU/NIOSH review in the 

iticality has ever occurred at 

RFP. The nature and extent of the criticality safety program was not reviewed 

SC&A: 

SC&A: 

. W

c

ORAU: 

us?

ORAU: 

workers, however, were 

SC&A: 

All the available information was reviewed. Since the original TBD was published, 

additional information about unusual occurrences has been discovered and will be 

cDowell-Boyer will be adding more information on incidents in the TBD as 

additional information is found. 

NIOSH:

SC&A: 

id Rocky Flats handle or process recycled thorium?

There is no evidence of recycled thorium. According to Section 2.4.3.1 of Chapter 2,

thorium was handled in limited quantities and had several different forms. 

17. What assessments of criticality safety and radiological control at Rocky Flats did 

preparation of the TBD?

ORAU: There is no evidence that an accidental nuclear cr

in the preparation of the Technical Basis Documents. 

Let’s defer this one as well until later so that Bob Alvarez can be on the call. 

18 ere there any incidents involving the furnace operations with respect to ignition of

ombustible gasses, flame flashbacks, etc.? 

Furnace operations at RFP have resulted in accidental airborne releases by a 

number of mechanisms. Rapid oxidation in a furnace in Building 707 (Jmodule) did result in an accidental airborne release.

SC&A: At RFP, there is a history of blowbacks. Can you characterize these blowouts for

We are aware of a blowback in J module in Building 707. There were vacuum

furnaces that did have contamination incidents. The

monitored and would not likely have exceeded the 10% guideline. 

19. How did NIOSH determine which incidents were to be included in the site profile?

ORAU: 

summarized. 

Laura M

We have reviewed the history of the incidents and have gone through them carefully.
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nts contains 27,000 pages of reports. We have gone 

through all of them. There are a few additional ones where pages are missing. It 

them out. Instead, it looks like 

the pages were missing from the originals. We are trying to track this further. We

have found nothing so far that is any thing like the 1957 and 1969 fires. There were 

ill forward the report to the DRs 

Was there any involvement of support workers?

ORA

SC&A: on for RFP, we feel that these

of such

n to look for

ividual’s file. If there are 27,000 pages, there may 

ts. So when you are doing an individual’s dose 

to augment the records for individuals from

s

ay 

olved in one of these groups. 

NIOSH: For internal dose, to be more claimant favorable, we do assume a constant chronic 

intake as long as they were receiving positive bioassay results.

SC&A: 

20. What records retrieval efforts were made in support of the site profile review? Were 

any classification issues encountered during site expert interviews and records retrieval 

efforts?

ORAU: Information was retrieved from the Denver National Archives, University of 

Colorado library, Front Range Community College Rocky Flats Reading Room, the 

site profile research database, the Kaiser-Hill RF records center, and the Rocky Flats 

Radiological Health group. No classification issues were encountered. 

ORAU: The document set on incide

does not appear that this was caused by not printing

some who entered areas where respirators were required without respiratory 

protection but when they realized it, they left quickly. They, however, did get nasal 

smears. 

NIOSH: They will put whatever they find into a report and w

and to SC&A. 

SC&A: 

U: There were cases of contamination in controlled areas. Reports say the public was 

not endangered. Generally, they did find contamination, but there was no undue 

exposure. There was one exception to this which involved a tour group that passed 

through a contaminated area and 15 people got 10-4

 mrem.

When you have a dose reconstruction process going

incident reports are important. If a worker’s file is lacking any mention

incidents, then incident reports become important. It is a trigger to begi

bioassay data if it is not in the ind

be a lot of information on inciden

reconstruction, it will be important to try

these 27,000 pages. 

NIOSH: We agree. If such data is found, it will be placed in the person’s file. If it affect

dose, ORAU is committed to go back and redo the dose reconstruction. Where the 

POC is near 50%, this will be even more important. 

ORAU: The reports do not identify names, but usually do identify groups. Therefore it m

be difficult to identify if a specific individual was inv
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SC&A: Th

rking on getting this information. 

 obtain worker input for the site profile? 

OR formally contacted. 

ngs.

ed by 

ORAU: 

SITE DES

C&A: ignificant

external or internal hazard. Why isn’t uranium a significant radionuclide with 

significant external and internal dose potential in the early days when RFP workers 

ORAU

e response is fine as is.

ORAU: Roger Falk is wo

SC&A:

21. What worker organizations were contacted to

Who are the contacts for these organizations? 

AU: No worker organizations were

SC&A: The response is fine. SC&A has minutes of these meeti

SC&A: 

22. A “job exposure matrix” was developed as part of a DOE-funded study perform

the University of Colorado Health Sciences Center and the Colorado Department of 

Public Health and Environment (and cited in ORAUT-TKBS-0011-6), which matches 

external dosimetry results with year-by-year building assignments and job titles. Has 

the matrix been obtained by NIOSH for use in DRs?

NIOSH: Brandt Ulsh has been working to see if they can be released to ORAU. 

We are still working on this too. 

CRIPTION (ORAUT-TKBS-0011-2)

S Page 7, Section 2.4.3.1, states that DU is not generally considered a s

performed a lot more operations with uranium (enriched, natural, and depleted?

: Let’s defer this until Bob Alvarez can be on the call. 
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ate: September 8, 2005 

articipants: 

NIOSH

OCCUPTI

SC&A:

to provide a revision to this site technical basis document? Is NIOSH currently revising 

nced

orated. 

t

t

2. Wh nce or absence of medical protocols used at 

ne the need for radiological exams during his tenure at RFP or prior to his 

All available document sources were searched for RFP protocol. A written protocol 

an, 

n

-rays were performed from 1952 through 1974, that chest x-rays 

DR. FURM e

ed 

for rad workers,

RFP Occupational Medical Dose TBD Conference Call Summary

D

Time: 10:00 am - 11:00 am EDT 

P

ORAU: Robert Meyer, Craig Little, Joe Furman, and Theresa Lopex 

: Brandt Ulsh, Greg Macievic 

SC&A: Harry Pettengill, Tom Bell 

ONAL MEDICAL DOSE (ORAUT-TKBS-0011-3) 

1. The above listed document is dated February 2004. It is noted that a lot of cross 

referencing is relied upon between various site profiles. What is NIOSH currently doing 

any of the TBD documents for other sites that would require subsequent changes to this 

document?

ORAU: The Rocky Flats TBD has been reviewed for consistency with all cross-refere

documents, and when we do the update in December 2005, this will be incorp

We are ensuring that all TBD updates for all sites will be updated to refer to the lates

version of ORAU-OTIB-0006. We will also review the SC&A issues and ensure that 

these are resolved and included in the latest update of the TBD. We will ensure tha

we use the most up to date references as we issue updates to TBDs. 

SC&A: 

at does Dr. Furman know about the existe

RFP to determi

time at RFP? 

ORAU: 

was not located but it is the recollection of former employees, including Dr. Furm

that the protocol would have been consistent with other DOE sites. We are certai

that lumbar spine x

were performed annually through 1984 for most workers, and that after the mid1980s, chest x-rays were performed less frequently and only for certain classes of 

workers, as described in Section 3. These conclusions were drawn from reviewing 

worker files as well as the recollection of employees in the medical department.

AN: By radiological exams, do they mean lung counts? If so, I don’t remember th

frequency (yearly for rad workers?) but someone like Steve Baker, who perform

the examinations, would know. If they mean medical examinations
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SC&A: 

Dr. FURM

DR. FURM

do a chest x-ray with each exam. Until the mid-1970s,from limited records, 

it appears that chest x-rays were done sporadically. Most annual chest x-rays were 

C&A: Were annual spines done separately from medical exams during the period from 1950 

DR. FURM

DR. FURM

ers that presented themselves at the clinic and 

requested an x-ray. If they chose not to come in, then they would not have had a 

C&A: We notice that you make reference to Steve Baker in your response. Was he the one 

R. FURMAN: Yes. 

SC&A: 

han

ORAU: he TBD assumes that 

all workers received an annual chest x-ray through 1984; lumbar spine x-rays (one 

est x-rays for certain

workers from 1985 to the present (see Table 3.2-1). Food handlers and nurses, 

of lumbar spine x-rays and an annual chest x-ray. 

they were required to be yearly, and since 1984 when I first arrived, and even 

somewhat before, were reliably done on a yearly basis. 

I should clarify that this question was related to only chest x-rays and does not relate 

to lung-counts for radiation workers. Also, these are performed annually. 

AN: What we have heard about the possible use of PFGs at RFP seems to be 

consistent with other sites in terms of protocols?

AN: There were no formal or written protocols until the mid-1980s. We did, 

however,

not done by protocol but by tradition. Routine spines were done in the early years as

part of the pre–employment medical examination. 

S

to, say, 1975?

AN: No, they were done at the time of employment. No information was found to

suggest annual spine exams were performed. 

SC&A: Were chest x-ray done up into the mid-1980s as part of the annual medical?

AN: Yes, but in the early days this may not have been done for everyone. We did 

medical examinations on those work

medical examination every year. 

S

who did the whole-body counts (WBC)?

D

3. Were x-rays used routinely as part of medical certification for respirator users, food 

dlers, nurses, etc? 

Specific historical protocols have not been located. Therefore, t

set) for all workers hired between 1952 and 1974; and ch

therefore, would not have necessarily received x-rays if they were employed after 

1985 to the present. Prior to that time, the TBD assumes that they would have 

received one set
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er than just annually – wasn’t that by 

President Carter in 1978? We “formalized” the rules for x-ray frequency in the mid -

nt 

for periodic chest x-rays since TB became less of a public 

health problem. I don’t know when that was – sometime in the late fifties or early 

been getting annual chest x-rays 

through the 1970s, possibly to the mid 1980s, but no later unless they were an 

rker who had worked in asbestos at least 10 years or were a beryllium

worker. 

If a claimant got an annual medical examination, it is assumed that the claimant also 

until

-ray

ed

C&A: From personal experience of folks at SC&A, we realize that respirator users received 

nd sign off

on their ability to wear a respirator. 

DR. FURM

C&A: But isn’t it known that the health and safety department folks had their own tracking 

DR. FURM re in

ions.

C&A: You are correct that regulations began to be enforced after the Executive Order, but 

ons

C&A: What is known about routine x-rays for food handlers?

DR. FURM

DR. FURMAN: I believe we had presented the required frequency? An Executive Order was 

published to limit x-rays to those necessary, rath

1980s I believe, reducing it to every 5 years if they didn’t have a specific requireme

such as annually for beryllium workers. Food handlers and nurses were not 

considered to have a need

sixties at the latest. Respirator users may have

asbestos wo

got a chest x-ray. Getting annual medical examinations, however, was spotty up

1979. 

NIOSH: In doing dose reconstruction of the claimants, we assume that they got a chest x

when they had their annual medical examination. Since many may not have receiv

an annual medical examination in the early days, we are likely overestimating the 

dose from annual chest x-rays, but we feel this is claimant favorable. 

S

x-ray examinations in order to receive medical certification to wear the respirator. 

The health and safety department sent folks to medical to get chest x-ray a

AN: We did certify respirator wearers, but we did this at the time of their annual 

examinations. 

S

system to assure radiation workers had necessary certifications? 

AN: That might have been so in the early days, but once medical protocols we

place in the late 1980s, it was done as part of the annual medical examinat

S

implementation was very slow. The Executive Order also provided specific dose 

guidelines for diagnostic x-ray procedures for the first time. Also, these regulati

were not consistently applied at DOE sites until the mid-1980s. 

S

AN: Routine x-rays for food handlers were not documented in RFP records. We

also checked with the State and they did not have a mandatory requirement that we 

had to do these. 
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SC&A: 

DR. FURM

DR. FURM

SC&A: 

4. De

eva

ORAU: 

ationally required x-rays (i.e. – annual 

exams, pre-employment physicals, and exit physicals). It does not include medical 

C&A:

The

TDB does not address this. 

NIOSH: 

e. Dose

calculations in the TBD are based on the assumption that there is one prere.

SC&A: 

e

NIOSH: n nonto

nd

would need an x-ray. This is taken into account in the POC computation in IREP. 

In regard to Beryllium workers, we know they were supposed to get an annual chest 

x-ray, what is the period of time that occurred? 

AN: The first Beryllium case we looked at was in 1984. NIOSH came in quickly 

and ensured that we enforced their standard and that annual exams were more 

complete in their scope. 

SC&A: Did that become part of their annual physical?

AN: Yes. 

fine what is meant by occupational medical radiation exposure? Does it include 

luation of foreign material resulting from machining and milling accidents?

Occupational medical dose is a default estimate of the radiation dose that was 

potentially received by workers from occup

x-rays arising from injury, accidents, or treatment, as these would vary greatly by

individual and should be noted in medical records. 

DR. FURMAN: Is that not the exposure from x-rays taken as a result of an occupational injury? 

I never heard anyone refer to any other work-site exposure as “medical”. 

S A lot of accidents and incidents that happened at RFP were due, in part, to the fact

that the facility was a large metallurgical facility and workers would need often to be

evaluated for the presence of foreign particles or splinters (e.g., Pu) in wounds.

You are correct. This, however, is a generic problem. X-rays resulting from work

are part of occupational exposures. They are often just included in worker 

background statistics and are not part of the occupational medical dos

employment and one annual examination only, as part of the medical exposu

There is another area where the general rule may not be appropriate for all workers 

(such as machinists who were in higher number at RFP). Machinists may have 

received varying levels of exposure. How are you dealing with those who are outsid

the general rule? 

It is handled in the statistics of background exposures for the site. If you were i

nuclear work classification, a certain percentage of workers are known, in general,

get hurt on the job (whether you were a nuclear worker or a non nuclear worker) a
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 types of x-ray exams. 

That is different and sporadic and is handled differently in developing the individual 

That kind of data is not available unless it is in the individual worker's personal file. 

These become part of the statistics of injuries at a site. 

SC& :

seen that it is handled differently than you describe, based on what we have seen at 

at is

and

r was doing when he was working in a

nuclear facility. When we looked at individual cases, using Ron Kathren’s ORAUis not

s is new information that you have just described and I am not 

aware of this. Please explain this strategy. 

IOSH: You said it correctly. If I said anything above that is contrary to what you have 

stated, I did not mean to. I agree with what you said. 

SC& :

s.

 

When working at a job, such as doing machining, any x-rays associated with that 

IOSH: But at RFP, most of what we are talking about for these kinds of injuries do require 

have a Pu splinter, we

would be interested in this if the claimant has skin cancer, but this would not be of 

C&A: NIOSH has gone a long way in assessing dose from x-ray exposures of pret

y were machining and got a Pu 

splinter, it seems this should be treated differently. This should be looked at, since 

-

ave chest x-rays for them to be cleared to go back to 

Just based on the number of people entering a site, you can statistically evaluate the 

number of workers that will likely be required to have certain

claimant’s POC. You can't look at every nuclear worker who might get a splinter. 

A We have already gone through a lot of discussion on this issue but I think we have 

other sites. Any individual x-ray, for instance if you trip and fall at home and need an 

x-ray, is not considered a part of the worker’s occupational radiation exposure th

used in the input to IREP. Such a dose was something that was just off the table

not considered to be related to what a worke

OTIB-0006, background radiation from these other sources are not included. It

realistic to do this. Statistically when you get an x-ray like this, it is not a part of the 

overall process. Thi

N

A To restate then, from what we have seen, any additional x-ray like the example above

is not part of the IREP and is beyond what is used to include annual chest x-ray

injury are not captured and used for entry into IREP. Such injuries are handled the 

same for any machining occupation, regardless of the type of facility involved.

N

x-rays. If the claimant’s individual file showed that he did

interest for other cancers. So only a few cancers would be involved. 

S

employment and annuals. These x-rays are important, but they are not unique to jus

nuclear industries. No one is suggesting that x-rays incidental to work that are seen in

off-site hospitals should be included. But when the

the exposures are at varying levels and could be significant. We note that ORAU

OTIB-0006 Rev. 3 also stresses the importance of including the dose from chest 

x-rays done during termination exams. The TBD does not include these in the 

definition of occupational medical examinations. But the workers, as part of their 

jobs, were also required to h
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TIB, but is not included in the TBD,

nor is it covered sufficiently in the OCAS 2002a guidelines. 

IOSH: We will have to give this entire issue further thought as your points are recognized. 

SC&A:

5. Were workers required to have radiography exams as part of a return to work clearance?

ORAU: ite

rtment did not do any procedures to 

document such x-rays. No radiography was associated with this return to work 

SC&A: 

n

Dr. FURM

SC&A: 

6. The

NIO

exceeded 25

IOSH: have not seen any high doses that came from

RAU: ee if there are any problems with x-ray systems. I would not 

ort

SC&A: 

work. This circumstance is also discussed in the

N

According to Dr. Furman, never. If there were fractures, they were referred to off-s

medical facilities and the RFP medical depa

process. Some could come in as a referral. 

If an injured worker was referred off-site for treatment and follow-up, did RFP 

receive diagnosis and prognosis information, including x-ray results, as part of retur

to work protocols?

AN: That’s possible, but it was not part of the worker’s RFP medical record. 

TBD suggests all radiography occurred in the diagnostic range of (30-250 kVp). Is 

SH aware of the acquisition of any x-ray units that were installed at RFP that 

0 kVp? 

ORAU: It is the recollection of Dr. Furman and other employees that the voltage of the x-ray 

units could not be set higher than 250 kVp. Maximum machine settings were 

presented in Table 3.3-2. 

SC&A: Was there any other x-ray generating equipment at RFP?

N Yes, but they were all locked up. We

the vaults. No one seems to remember that. 

O We will do a scan to s

expect this, since there were no written protocols. In one of the reports, part of rep

is missing and we will need to follow up on that. 

Rocky Flats Plant, being a Federal facility and being in the Agreement State of 

Colorado, would they have had strict procedures when equipment was brought in, for

example, an 192Ir NDT source? But this was probably provided and controlled by a

private off-site contactor, who may or may not have registered its source properly. 

How do you consider this?
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We will go back and review for this possibility..

SC& :

7. (by medical record 

or protocol documents) that show workers were not exposed to voluntary chest x-rays, 

ORAU: rs defining “occupationally required” x-rays, I don’t think 

that voluntary, non-mandatory x-rays should be included in the TBD regardless of 

DR. FURM inician

for having an “off schedule” chest x-ray done. For 

-ray

hat

review of many hundreds of charts. But remember, before the 1980s, 

-

were

take what was offered.

R. FURM you could say that a large number got annual medical examination 

ORAU: 

medical examinations. About 50% did get them in the early days. 

Some got them every couple of years. Generally the workers got a chest x-ray with 

their annual examination. Accreditations were done up though the 1980s. This was 

IOSH t treat non-radiation workers any differently than radiation workers for 

purposes of dose estimation for medical exposure. 

NIOSH:

A

Section 3.2 discusses exam frequency. Is there any physical evidence

even when policy did not require them?

Given the legal paramete

whether they may have occurred. 

AN: Well, I cannot say that at some time an employee did not convince a cl

that they had a valid reason

example, if an employee complained of a cough that had lasted a couple of months, 

perhaps the clinician would order an x-ray rather than just referring them to their 

private physician. As Medical Director, I discouraged that, since the DOE did not 

want us to compete with the private sector. I doubt that anyone would order an x

at the whim of the employee; that would be contrary to our training and ethics. T

would require a

there still was a tendency to just give everyone getting an examination “everything”, 

thinking we were doing them a favor. Providing examinations tailored to their work

place exposures or, if a voluntary examination, tailored to their age, did not start until 

the 1980s. At that time, with increased emphasis 

SC&A: We have read in the TBD that a lot of chest x-rays were voluntary.

DR. FURMAN: If they were offered the opportunity to receive an annual exam, then they

offered to everyone. They tended to

SC&A: Am I to take it that the TBD's dose calculation is reasonable and that all workers were

given an annual chest x-ray?

D AN: I don’t think

in the early days. 

From what we have seen, I would say that a majority of the workers did get their 

annual examinations which included a chest x-ray. But a lot of workers did not show 

up for their annual

not just done for rad workers, but for all employees. 

N : We don’
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8. Is there any physical evidence to show that special surveillance exams for beryllium,

asbestosis, and respiratory protection were always scheduled as a part of the annual 

rovided in the technical document?

C&A: Supervisors had to ensure tracking of that. In the early days you didn’t see any 

R. FURMAN: You’re correct. After 1986, it was tracked and the special surveillance was part 

to making sure people were getting their required annual medical 

examinations. 

NIOSH: ave gotten more dose assigned than 

they actually may have received because we assumed that everyone had an annual 

% possibly did not. Even with possible extra xray exams for Beryllium and asbestosis, the assumption that everyone got an annual 

and a pre-employment physical would still be claimant favorable. 

SC&

DR. FU special x-rays because they were, for 

example, a beryllium worker. Workers were sent to medical and medical did 

NIOSH: claimant had more than the routine medical x-rays would likely 

be recorded in the claimant's personal file. If extra exams are found in the file, the 

l medical x-ray 

exposure history. 

SC&A: 

9. Are there any records or studies to document a low retake rate alluded to in the technical 

SC&A: 

physical which is an assumption p

DR. FURMAN: In the 1960s and early 1970s there is evidence that an annual physical was not

always done. It appears that there was not tracking of this requirement except by the

individual supervisors. 

S

tracking process, but after 1984 as the Federal regulations were implemented and the 

guidelines started to be followed in 1985 and 1986, then you did start to see 

compliance in a big hurry. 

D

of the annual exam. Bob Bistline got the computer tracking started and gets a lot of 

credit for having done so. This prevailed through the mid-1980s. Beryllium 

propelled us in

Prior to the mid-1980s, claimants would they h

physical, when in fact as many as 50

A: Is there any evidence of an increase in chest x-ray procedures other than those 

required on an annual examination?

RMAN: You never had to send them back for any

everything. They would do comprehensive physicals when workers were set up for 

annual physicals. 

Any evidence that a

DR will take that into consideration and will add the dose into his tota

document?
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ere is data indicating a low

re-take rate.{Note: I believe that the Complex-wide TBD values assume a retake rate 

DR. FURM

al

could be a lot higher and may not be reasonably accurate for medical facilities at 

ORAU: 

rates were assumed to be low in the post1980s time frame but we don’t have documented information for the earlier period. 

NIOSH: tribution with a 30% uncertainty assigned (root 

mean square of uncertainty). Retake uncertainty is just one of the five or six 

SC&A: might 

IOSH: We will look at that and follow-up 

SC&A: 

re there any records or protocols to demonstrate that radiological processing equipment 

was properly maintained and serviced to reduce and minimize worker exposure? What 

RAU: Inspection records of the x-ray machines were kept by Radiological Control. These 

were kept by Medical since the 1980s. Finding them

would be difficult.

ORAU: 

e documentation of maintenance 

ORAU: Prior to the 1980s, there is little information. Post-1980, th

– I will confirm this) 

AN: I found no data on that. 

SC&A: ORAU-OTIB-0006 makes reference to rigorous studies performed in large medic

research facilities under controlled conditions. It assumes retake rates of 3%. This 

RFP. 

The retake rate is assumed to be 3%, but this in not significant enough to incorporate

separately and they just dropped it. Retake

We will look further into this. 

ORAU-OTIB-0006 uses a normal dis

uncertainties. 

We still think you may want to take a more rigorous look at retake rates, as you

find they are low, especially for spine x-rays. 

N

10. A

records exist for the purchase and maintenance of all x-ray equipment, radiographic 

processors and supplies for the years 1953–1989? 

O

records are now difficult, if not impossible, to find. The records of outside agency 

inspections and maintenance

DR. FURMAN: Is it true to say that no one could come up with the hard copy?

We looked. The State said they did not have them. It is likely a private contractor 

that did this and it was handled as consults. A member of our team spent a week in 

Dr. Biseline’s office and could not find any medical protocol information. 

SC&A: This is not inconsistent with other sites. There is littl

or associated records of machine outputs and calibrations. 
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e and did examine equipment. 

RAU: We have a look-up table in ORAU-OTIB-0006 that deals with uncertainty. The 

SC&A: 

these additional issues. 

the justification of the 

30% uncertainty assumption and the retake rate of 3%. We will follow up on this. 

SC&A should follow up on this too. 

SC&

s,

temperature, and quality of the chemicals need to be evaluated.. 

IOSH: We agree. We will get back to you with an evaluation of the uncertainties in ORAUC&A:

11. Did

util at was the make and model?

What was the period of use? How many exams were performed?

RAU: X-ray equipment that was present on-site has been listed in Table 3.3-1. We have no 

DR. FURM

n or evidence of use in the many charts I had occasion to review over the 

.

much. 

C&A: Are there no records of how much film was bought?

DR. FURMAN: State did come on-sit

SC&A: My point in this question is that many things affect dose beyond the machine output. 

Principal among these is the processing equipment to make the point that a number of 

aspects have a bearing on uncertainty. 

O

general uncertainty is assumed to be 30%, or one sigma. 

With all the factors that have a play on uncertainty, there should probably be a review

and rethink of the validity of the 30%. More needs to be known about the physics of

the x-ray unit and the protocols used at the time of the exam. It seems important to 

come to grips with

NIOSH: We agree. We need to look at ORAU-OTIB-0006 and review

A: If you look at it closely, there is a lot of good information on all types of variables 

that can affect uncertainty in ICRP 34 and NCRP 102. This, however, is early 

vintage stuff. Even though the dosimetry is good, it still is based on the use of 

optimized data. Such factors as filtration, voltage output, kVp, buckys and screen

N

OTIB-0006. 

S

the RFP site acquire and use portable x-ray units? Can NIOSH better describe the 

ization of photofluorography (PFG) units at RFP? Wh

O

additional data at this time. 

AN: Portable medical x-ray units were not used at Rocky Flats. No record has been

found indicating how or even if the photofluorography machine was used. I saw no

mentio

years. Well into the 1960s we saw the typical 14 x 17 films for the usual x-rays, but I 

have never seen the smaller PFG x-rays films in the medical records I have reviewed

The PFG unit would have been in the medical department, but it was not used

S
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ORAU: val of the PFG unit in the 1968 (see Table 3.5-2), page 21. 

It was there in the medical department from 1953 to 1968. But by assuming one PFG 

SC&A: ate per each PFG exam in ORAU-OTIB-0006 is a good and 

reasonable assumption. A lot of PFGs were placed in vans and became portable 

increase is a valid way

to deal with this. 

NIOSH:

C&A:

12. Many radiographic procedures are performed offsite on a medical referral basis. Does 

wor

wou se.

DR. FURM all as just an injury or illness. It was all treated the same. 

85’s

ould

have been only for former worker follow-up exams. 

Discussion

C&A: Slide on page 26. There is still a need for NIOSH to better define its interpretation of 

what constitutes occupational medical exposure. NIOSH should update OCAS 2002a 

guidelines to incorporate what we discussed today. Clearly, the medical exposures 

from the pre-1985s may be a problem, and the TBD does not cover this uncertainty 

well. 

SC&A: Slide on page 27. The TBD does not address the potential for sealed sources being 

used in the medial clinic and does not catalog the number and types of x-ray 

equipment available for use and the conditions of use. 

DR. FURMAN: No evidence of that. 

We saw and reported remo

exam with each annual physical, this results in a huge overestimate when most 

workers probably did not have PFG exams. 

The 3.0 rem estim

screening units on mobile platforms. In the TBD, we note that Ron Kathren got 1.5 

rem in his calculations for the dose delivered by each PFG examination and he then 

doubled it to 3.0 in order to be claimant favorable. This 100%

There is no evidence whether the PFG was used or not used. 

S

NIOSH have evidence to determine the amount of offsite radiographs performed on 

kers at the request of the employer and as a condition of employment? An example 

ld be follow-up exposure to evaluate the presence of chronic beryllium disea

AN: We treated these

SC&A: Is there a possibility that pre-employment exams were done off-site in the pre-19

era 

DR. FURMAN: There was no evidence that any of this was contracted out up until the year 

2000. In 2000 it was all contracted out. Prior to that, if it was contracted out it w

of the PowerPoint Slides 

S
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ORAU: ere

made and how the equipment was maintained. We have no documents on protocols, 

ocedures.

C&A: I believe the point we are making is variability is the rule, rather than the exception. 

u may not want to rely on optimized conditions on protocols and procedures to 

ere is great variability directed 

DR tor preferred. Some doctors 

like the x-ray dark. With no rigid procedures early on, if a physician wanted to vary 

t

potential dose. Sometimes this was five to six times that of a conventional x-ray. 

IO

&A: m and 1 PFG exam per year, that is a claimantSC&A:

NIOSH ut

SC&A: uld only emphasize that there is a need to go beyond the specific output of the 

d consider all these other potential sources of uncertainty. We appreciate your 

lize that some exposures are hard to get one’s arms around it. 

We have been hard pressed to come up with documents of when purchases w

processes and pr

S

Yo

make estimates when no record exists. For example, th

by the physician. 

. FURMAN: Yes, that is true and it also involved what the doc

the output a little bit he just did it. It depended on the exam. 

SC&A: Slide on page 28. RFP did have the PFG unit there in the medical department for 

nearly 20 years, so the potential existed to use it. These old PFG units had the highes

N SH: Yes, the references do grant that. 

SC If you assume 3 rem per PFG exa

favorable way to deal with the PFG potential dose. In regards to the retake rate of 3%

for routine x-rays post-1968 and spine x-rays, NIOSH needs to look at that. 

However, it's probably does not need to be increased more than 100% as Ron Kathren

did with PFG exams. 

Slide on page 29. There is a lot of dose impact just due to optimizing the use of 

bucky systems. It is likely, however, that this information no longer exists. 

: We will have to look at the ORAU-OTIB-0006 again to see if it does properly lay o

these errors and uncertainties. We will have to go back and see how they did it. 

We wo

tube an

response and rea
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R

Date: Sep

Time: 3:00

Participant

ORAU Little, Laura McDowell-Boyer 

al Participants: Amy Dean (Scribe), Advisory Board (Mike Gibson) 

O

SC&A

1. 

 

tions and availability of data and SC&A gives 

ins.

se

tim

RF

195

03 pad

965 glovebox fire

ORAU

ebox fire was not as significant as 

SC&A:

FP Occupational Environmental Dose TBD Conference Call Summary

tember 6, 2005 

pm - 4:00 pm EDT 

s:

: Robert Meyer, Craig

NIOSH: Brandt Ulsh, Greg Macievic, Sam Glover 

SC&A: John Mauro, Joe Fitzgerald, Abe Zeitoun, Tom Bell 

Addition

CCUPATIONAL ENVIRONMENTAL DOSE (ORAUT-TKBS-0011-4) QUESTIONS

:

Could NIOSH provide a time line, which makes a clear distinction between operations, 

data availability, and types of data?

There is a lot of good information on opera

ORAU and the authors of 0011-4 a lot of credit for what has been done. On the other 

hand, it would be helpful to have a timeline that describes and distinguishes the types of 

operations and what data is available during both Phases I and II. The TBD has no 

references or discussions that help the reader to better understand when Phase II beg

There are the three major accidents. Then there are some 24 to 25 discrete events 

between 1964 and 1965. The matrix and the interaction of these and the relevance the

events have been summarized in the ChemRisk 1992 report. Figure 6-1 has a nice 

eline but lacks information on the phases. Five major accidents were identified for 

P. These were the:

7 fire

9

1

1969 fire 

1974 controlled value failure 

: When you are looking at the ChemRisk report, this is actually Phase I. Phase II was 

done in more depth in later reports. The 1965 glov

far as release to the environment was concerned. We have the document and we will 

deal with this in the case of claimants who were involved. 

It seems easier to find out what was happening during Phase 1. But in Phase II, the 

studies done do not seem to make it clear which phase you are reading about. 
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ORAU e

perations. The Rope et al. document (1999) deals with Phase 

II, as does the Rood and Grogan 1999 document, and these are helpful in 

BD

SC&A: or instance on page 8 in Section 4.2.1,

the TBD does not refer to what phase. It is important that this is made clear to the 

SC&A 

2. Table 4-1 of the TBD presents the estimated 50 percentile and 95th percentile annual 

ed

on reconstructed radionuclide source terms and atmospheric dispersion models 

employing Monte Carlo techniques. For the purpose of dose reconstruction, is NIOSH 

ilarly, in 

Table 4-2, does NIOSH recommend that the average or maximum estimates of annual 

intakes be used for the years 1965–2002? 

In regard to Pu intakes from 1953 to 1964, it is not clear what values are being used and 

e reasons for choosing either the 50th or the 95th percentile. As noted in the question, in 

Table 4.2 on page 17, it is not clear if NIOSH is recommending that the average or 

NIOSH

H uses the geometric mean of the lognormal distribution 

with a geometric SD of 3.0. Usually we use the 50th percentile. Of course, if the 

SC&A

3. 

ould NIOSH describe the rationale for 

applying these different parameters for particle size? 

The s,

you

rea change

fro d

: John Till’s report was for Phase II, so you know when you are in that report you ar

dealing with Phase II o

understanding Phase II operations. Both these documents are referenced in the T

reference list. But SC&A makes a good point and we will do a better job to make 

sure that this transition between phases along with timelines is more evident during 

revisions to the TBD. 

The authors do no seem to refer to Phase II. F

reader so that he knows which phase they were a part of. 

th

plutonium intakes from 1953 to 1964. It is our understanding that these values are bas

recommending that the 50th percentile or the 95th percentile values be used? Sim

th

maximum estimates of annual intakes are to be used for the years 1965 and 2002. 

: Task 5 may be better able to address this. When we are overestimating dose in a 

claimant dose reconstruction, we use the 95th percentile in most cases. Normally our 

response would be that NIOS

claimant has bioassay data in his record we use that. 

:

What is the relationship between the 15μm aerodynamic equivalent diameter (AED) and 

the 5μm activity median aerodynamic diameter (AMAD)? Why is it that the predictions

and models dealt with particles < 30μm AED? W

TBD appears to move from 5 um to 15 um in different places. If you go the graph

see the use of 30 um. SC&A would like to see more commonality for whoever is 

ding document, so that it is clear to the reader when there is a transition or a

m one diameter to another. SC&A does not disagree with what is done; it is just har
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to f

dia

we have to figure what fraction is relative to this. The ICRP AMAD of 5 

is only applicable for about 15% of the cases whereas an AMAD larger than 15 um is 

SC&A:

se 1 um. 

ORAU: 

of

indblown particulates that are larger than normally 

respirable particles that are traditionally included in dose assessments? What is the basis 

y

e

ental factors. 

ORAU

n of

had deposited on surfaces and was later transferred to the 

ollow. A transition is needed to allow the reader to better understand why various 

meters are being used. 

ORAU: We will go back and try to make these changes more apparent. When using the

aerodynamic equivalent diameter (AED), the highest value used is 15 um. But with 

the AMAD,

more appropriate in 80% of the cases. For instance, in the 1969 fire it is possible they

do not know the AMAD – it might be 1 um. 

Normal PuO2 behavior may not the behavior of the PU after the fire. If the physics 

and solubility is changing, then the stay time of Pu inside the body may be longer that 

they thought. Later as more Pu changes to Am, it stays even longer and a matrix of

changes is occurring. Another issue is when to u

NIOSH: This will be discussed at length in the internal dosimetry discussion tomorrow, so

let’s defer this until tomorrow. 

SC&A: The results of tomorrow’s discussion should be considered in assessing 

environmental dose. 

You may not even know what the solubility is. 

SC&A: 

4. Why does the occupational environmental TBD ignore the potential for ingestion

contaminated soils and other w

for not including inadvertent ingestion of radioactively contaminated soil and other finel

dispersed radioactive materials?

Ingestion needs to be taken into account in the equation of the dose. Environmental dos

may not contribute the higher portion of dose, but the reliability of what portion of the 

dose that comes from ingestion is important for the reader to understand as one of the 

possible contributors. These should be taken into account in dose reconstruction. 

Somehow the TBD should deal with how much internal intake is from ingestion and how

much is from inhalation, how much will stay in body in the various organ systems, and 

the proper dose for those environm

: When we started writing the RFP TBD, we were not looking at potential dose from

ingestion originally. But OTIB-0009 was later developed by NIOSH and this TIB 

does deal with ingestion from soil contamination. I looked at their method used in 

OTIB-0009 and found that there were two modes. One was from airborne 

contaminants deposited on food or drink and the second was from resuspensio

airborne contaminants that
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taminants. NIOSH is using this default assumption in revisions to the Bethlehem

Steel TBD. 

ORAU: 

mends that if the 

person is digging in the soil and transferring contaminationed soil to their face that 

.

NIOSH: 

SC&A: e

is

ngoing dialogue with NIOSH and is a more 

generic issue that SC&A has been talking to NIOSH about. 

SC&A: 

5. For

resu

The u,

but

C&A has recently discovered a report entitled “The Final Buffer Zone Sampling and 

Analysis Plan of 2002” which we will be glad to share with NIOSH. Appendix B deals 

lso looked at Sr. The U values were not really that high. The Cs 

maximum result was 2,830 Ci/g. 

soil contamination in the industrial area and in the buffer area. If these reports give us a 

worker’s face. When they look at the 50th percentile dose, only 0.1% of the dose was 

from soil contamination.

SC&A: Do you use 100 mg/day as the default assumption for the potential ingestion of 

con

We don’t know what they would use in OTIB-0009. I just realized that OTIB-0009

was applicable. I did look at EPA for more information on potential daily ingestion 

of contaminants. There are only a few studies on this; EPA recom

50 mg/year be used. The 903e area is the area with the highest contamination at RFP

But it didn’t make sense to use 50 mg/year. They did not define the mode of entry

and their analysis was not based on actual data. 

ORAU has done an analysis in order to attempt to address this issue and will put 

something in the TBD to close the loop on this issue. 

We would like to see why you do not use 100 mg/day. I remember that some of th

assumptions in the TBD were low, so there may be merit in reviewing the use of th

default assumption. This has been an o

the pre-1965 time period, has NIOSH taken into account that the continuous 

spension of Pu and Am?

re appear to be two dimensions here. We have discussed the particle size of the P

should also consider the physical dimension. 

S

not with the industrial area, but instead with the buffer zone outside the industrial area. 

The report deals with a lot of calculations of the contamination in soil. Appendix E 

shows that for 238Pu monitoring, they collected 656 samples. The maximum result was 

2,610 pCi/gm. They a

Also SC&A understands there are “DOE Close Out Reports” that include good data on 

good idea of the levels of contamination, then resuspension can be better understood as 

well as the doses that are occurring. This is an important issue as SC&A believes this can

lead to high numbers. DOE collected all the samples and made a deal with EPA 

regarding the use of this data. 
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NIOSH .

SC&A: f

on.

ize fraction) were released from the 903 pad over several 

users, mainly as a result of mechanical disturbance and wind action (Weber et al. 

ORAU: vent. It does not take into account 

all of the resuspension that occurred later. The page 14 statement comes from the 

John Till RAC report, which discusses this. In the Till report, they were able to 

capture data down field and were able to figure resuspension during all high wind 

events, so that the resuspension estimates become pretty awful. Those estimates were 

compared with actual Pu release and corrected. 3.1 Ci was dominated by 9 events, 

which were observed and quantified on the 903 pad. Americium was taken into 

account and the activity shows about 30% of Am in the airborne Pu releases. 

SC&A:

6. Has NIOSH evaluated the degree to which RATCHET can be used to reliably simulate 

the atmospheric dispersion factors onsite, in the immediate vicinity of the releases, which 

is characterized by several closely constructed buildings? Wouldn’t these building wakes 

prejudice the results?

SC&A has a concern that RATCHET (an offsite model) may not be accurate enough for 

areas with buildings. Has ORAU made any adjustments to account for onsite building?

The second point – if used during period with no data prior to the early 1960s, Figure 4-

A-3, why is there no data goes beyond the 50th percentile – you would think that the 95th

percentile should have been modeled too. In 1969 the 50th is measured. If you compare 

that with the 1957, it is 10x higher in 1969 than in 1957. We realize the RATCHET code 

: We have been using air samples data and some of the particles will be re-suspended

Therefore air data is better than soil contamination. 

In Table 4.2, it appears the intake from 1965 to 1972 releases are higher by a factor o

2. 

ORAU: The RATCHET code does include this. The resuspension factor was developed by 

them as their own factor. ORAU will go back and look at this more closely. In the 

Appendix of the TBD, it does mention that resuspension was used. Rood and Grogen 

et al. explained this and what factors they used for resuspensi

SC&A: There was an example of a waste treatment plant where the waste water was used to 

spread over the soil to control wind spread of contamination. The DOE close out 

report may cover this. The TBD on page 9 explain that an estimated 3.1 curies (50th

percentile) of Pu (<30 um s

1999). And yet on page 14 of the TBD according to Weber et al. (1999), the highest 

release estimate for 239Pu in the 903 storage area was on the order of 1000 g, which 

corresponds to 62 Ci. The discrepancy between these two statements needs to be 

explored. How much really did leak out of the drums and did ORAU assume that 

some of the PU was converted to Am. 

On page 9, 3.1 curies related to a discrete release e
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is the e 4.1 for

1957 of 3.3 pCi/year, it would seem you should multiple this by 10, so that the 3.3 

comes 33 pCi/year. This 10X factor should also be applied to the 

since there were no laws in the 1950s on this. Therefore the 

1952 to 1964 gaps should use the 10X value 

OR

estimates were used. If you keep adding factors of 10 on 

C&A: There will need to more follow up and explanations of these margins of conservatism

it

we have dealt with before. 

ause it had source terms for the RFP site and the results could

d a release from a stack that was 53 meters high. 

However others say that 47 meters is more accurate for the stack height. Even with 

cks of this height would tend to rule out that buildings which are 

not this high would have a big input. They are looking at the average concentrations

best we have. But if you use the average (50th percentile) value in Tabl

pCi/year for 1957 be

years before 1957 as well,

AU: A lot of this was modeled on the 903 pad releases which would skew results. ORAU

will have to look at what other claimant-favorable things were used in the 1957 fire,

and what other conservative

top of other factors of 10, the values can become unrealistic. 

S

in regard to the 1957 fire to see if the conservative assumptions are logical and then

will be possible to determine if the X10 is needed. How RATCHET applies is one

ORAU: We use RATCHET bec

be compared with the 30 um data. RATCHET was only used for the period of 1953 

to 1957. One of the major events ha

these variations, sta

and the average (50th percentile) is probably OK. 
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all Summary

eptember 7, 2005 

ime: 10:00 am- 12:30 pm EDT 

Particip

OR e, Roger Falk, Jan Johnson, Liz Brackett, Karen Jessen, 

SC&A:

Additiona

Introducto

We will let t

out a revise

C&A General Comments:

unstana Melo provided corrections to the titles of two tables in the previously provided 

int presentation on preliminary talking points for the RFP conference call. It was 

ointed out that on page 17 of that PowerPoint presentation (see Attachment 6), the titles of the 

tw b

Coeffic

he PowerPoint slides were done later and provide some perceptions as to where we are now as 

in ge

12 of the v

The talking

sues.

e response in the matrix table is OK or 

ther clarification. We may not need to discuss the questions if it is answered OK. 

R

Could n?

RFP Occupational Internal Dose TBD Conference C

Date: S

T

ants: 

AU: Robert Meyer, Craig Littl

Tim Vitkus, Al Wolff 

NIOSH: Brandt Ulsh, Sam Glover 

Joe Fitzgerald, Dunstana Melo, Tom Bell, Bob Alvarez, and Kathy Robertson 

DeMers 

l Participants: Amy Dean (Scribe), Muddy Sharkey, Advisory Board, Mark Griffin 

ry Comments: 

Joe Fitzgerald walk us through the questions as we did yesterday. Jim Langsted sen

d Section 6 on external dosimetry. 

S

D

PowerPo

p

o ta les should read “Intake Retention Fractions for Pu-239 Inhalation” instead of “Dose 

ients for Pu-239 Inhalation.”

T

to prelim ary issues that were done 3-4 weeks ago. The internal dose discussion begins on pa

iewgraphs.

points are where we are and will likely change as we discuss all these questions and 

is

Let’s hav Dunstana Melo comment as to whether the

needs fur

O AU General Comments:

we make sure that each new iteration of the ORAU matrix has an update date and versio
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OCCUPA

SC&A: 

part

it us

ORAU: -13 was presented in the Appendix as an example of a data report that the 

DR might encounter in a claimant’s file. In this example, the parts per million by 

dpm

cident in order to

interpret the Pu lung deposition from the lung count results, which were americium

some of the records in Attachment C of the internal dosimetry TBD that 

m

ana

RAU: Use of offsite laboratories to analyze bioassay samples started circa 1993, starting 

ed Quanterra, and, circa 1998, 

with General Engineering Laboratories, Inc, Charleston, SC. Some bioassay samples 

as backup to the primary lab. 

C&A:

3. How

of p

reco ility values are 

applied, and on what basis, when solubility is not known?

RAU: These issues are being addressed in a project technical information bulletin, ORAUTTIONAL INTERNAL DOSE (ORAUT-TKBS-0011-5) 

1. Page 84 of the internal dosimetry TBD shows a “Health Sciences Record.” On this 

icular record there is a notation of ppm Am. How was this value determined? How is

ed in dose reconstruction?

Figure C

weight of the americium-241 (ppm Am) was calculated and recorded for the fecal and 

nose smear samples. The value of the ppm Am is calculated by converting the

(activity) values of the Pu and Am to mass by dividing by the appropriate specific 

activity (for Pu, weighted for the relative isotopic abundance for weapons grade Pu). 

The ppm Am is then the ratio, [mass Am/ (mass Pu + mass Am)] x 1, 000,000. The 

ppm Am value was used in real time by the dosimetrist who was managing the case 

data for the worker who was involved in a possible inhalation in

measurements. The project DR may or may not decide to use this type of 

supplemental information 

SC&A: The response in the Matrix table is OK. 

SC&A: 

2. It was noted on

so e bioassay analysis was done offsite. What offsite vendors were involved in bioassay

lysis for RFP? For what time period? 

O

with IT Analytical Services, Richland, WA, later nam

were processed by other laboratories

SC&A: The response in the Matrix table is okay. 

S

do dose reconstructors identify uptakes of Super Type S and other insoluble forms 

lutonium, particularly prior to lung counting, beginning in 1964? How is the dose 

nstruction process modified for these cases? What default solub

O

OTIB-0049, which is currently in review. The implementation, when the TIB is 
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he RFP in the same manner as for all the other sites with 

plutonium inhalation exposures. In general, when the solubility type is not known, 

e. 

SC&A: are including dose from the highfired PuO2. Are you following the approach in OTIB-0029 that was used for Y-12? 

get a non compensation when determining the POC, where the 

Type S issue might be important, we will hold the claim until the OTIB is issued. But 

B should also be out in two

weeks. It covers high-fired PuO2, but not high-fired uranium oxides. It does discuss 

it, but we don’t think that high-fired PuO2 will affect the dose.

OR

long-term retention of uranium.

SC&A: urdens of DU?

w about RU

and its trace daughters? It’s hard to know how the processes have concentrated them. 

At Hanford, the bag workers got high doses of Np.

ns might have had RU. This is discussed as part of 

study done and published in the year 2000. 

C&A: But there is some data regarding ingots that indicates this could be as high as 2500 

re about 

igh-fired oxides?

NIOSH: e dose that is not currently in the TBD. 

approved, will apply to t

the DR will use the type which results in the most claimant-favorable outcom

We are interested in learning more about how you

If you consider the high-fired PuO2 as Type S, we agree that is acceptable for all 

cancers except for lung cancers. What solubility class are you using for high-fired 

PuO2? 

NIOSH: For cancers other than lung cancers, Type M is appropriate. A new OTIB is in review

on high-fired oxides that has been sent to NIOSH for review, and it should be 

available soon. If you

very few are being held for this consideration. The OTI

AU: We focused on the Pu in the OTIB. In our experience, they have not observed any 

Then you are saying that there are no claimants with lung b

ORAU: For Pu, we see workers who have Am showing up in their lung counts and this is

stable over 30–40 years. 

SC&A: Yes, Pu does stay in the body for a very long time. But U does not. Ho

ORAU: At RFP, we did not have a fission product problem. 

SC&A: Did RFP process recycled uranium?

ORAU: Only 2 tons out of approx 8,000 to

S

metric tons and they did find fission products. It would be useful to know mo

this. Yesterday it was agreed this would be addressed perhaps as a broader issue. 

Does OTIB-0049 address any potential for h

ORAU: Yes, it will. If it can’t be determined, the DR should make that assumption. 

There will be guidelines on how to determin
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There is currently a TIB on how to fill in for missed dose, but there are also

guidelines for each site (not formal documents). These informal guidelines are 

IOSH: Filling in missed dose comes from a compendium of information. Task 5 comes up 

SC&A: t

s until the 

IOSH: These questions won’t be clearly addressed until later. 

C&A:

4. Wa s?

RAU: None that we know of. High ventilation rates in the processes and other populated 

they were not there.

SC&A: 

NIOSH:

RAU: We are not acquainted with that. 

C&A:

5. 

assessment calculations. What is the default assumption for the form of tritium? Were 

ORAU: ed from tritium gas released to 

the atmosphere. There is no evidence of other forms of tritium at RFP which would 

constitute a significant source of exposure. 

SC&A: What is the document? 

ORAU: 

written and distributed to the dose reconstructors. 

N

with specifics as to how do dose reconstruction. We will be glad to provide them if 

you have not seen them. 

Are you using OTIB-0029 that you developed for Y-12 for internal dose estimates a

RFP, or are you holding all but minimum/maximum compensation decision

OTIB 49 comes out? 

N

S

s any potential exposure from radon and/or thoron? If so, under what circumstance

O

areas would have precluded buildup of radon and thoron. 

SC&A: The response provided by ORAU addresses our question, but we are seeing lead 

peaks on lung counts. Were any workers monitored for 24Na?

ORAU: We did not look for fission products since we feel

NIOSH: 226Ra would not be there. 

Did it occur naturally and is that why we are seeing lead peaks?

Radon is an issue with Gravel Gertie’s, but this was not so at RFP. 

O

S

Section 5.3.1.6, Tritium, did not mention the form of tritium assumed in the dose 

there other forms of tritium present at the site? 

The default assumption is tritiated water, HTO, form
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C&A:

6. Wh d for assignment of internal dose to unmonitored or inadequately 

monitored individuals?

ORAU: 

ssay may also be 

assigned internal dose based on IBMA runs if it is needed to refine the estimate or 

SC&A: have an unmonitored individual, are you defaulting to OTIB-0002? 

th

estimate of dose. We might use OTIB-0014 if it 

involves environmental dose reconstruction. 

SC&A: 

IOSH: If you have a secretary or cook, you might give them an environmental internal dose, 

C&A: Will you be mentioning OTIB-0049 in the TBD? 

ORAU: sible to have all OTIBs mentioned in the TBDs. They are coming out all 

the time and may come out after we issue a revision of the TBD. The TBD is 

fic

that is specific to that issue at Hanford. It is not 

applicable to other sites.

SC&A: 

U

ORAU: rd of intakes by RFP workers to pure forms of these radionuclides. 

Pu is included in the weapons grade plutonium mixture. Contaminants in recycled 

SC&A: The response in the Matrix table is OK. 

S

at method was use

The DR has generic procedures and instructions to deal with these circumstances. As

appropriate, internal dose is assigned using claimant-favorable methods provided in 

OTIB-2, OTIB-14, and OTIB-18. EEs with specific partial bioa

determine dose from positive bioassay results. Maximizing or minimizing 

assumptions are made depending on whether the case is compensable or noncompensable. 

If you

ORAU: If the claim is currently in processing, we would use OTIB-0002 or OTIB 0018, bo

of which would result in an over

Is the environmental dose what you would consider to be that which occurs outside?

N

even though they were inside and more in a sheltered setting. 

S

It is not pos

designed to provide the basic data. For instance, some TIBs are directed to a speci

site such as what was discussed in the Hanford TBD. In the early 1943 – 1945, there

is a special TIB that is cited for use

7. There is no direction for dose reconstruction associated with neptunium, thorium, curium, 233 and 238

Pu. Has the potential for intakes from these radionuclides been considered?

There is no reco

238

U at RFP were not a significant internal dose issue.
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SC&A: 

ORAU: 

SC&A: il 1985, and things widely

varied from site to site. We feel there is a commitment to provide a strong rationale 

for some of the potentially dominant radioisotopes. The TBD seems not to address 

these. We also wonder what you are doing about dose for Am. 

ORAU en it is

r this. The urine program started in 1964, as discussed in 

Appendix A. We did not have a program to monitor for curium, but if we suspected 

C&A: How do you deal with the purification issue with Am, and can you be sure there is 

Am decays at rate of 20 ppm per month. With the recently purified Pu, and if it was 

less than 200 ppm, then that person was recounted once a quarter for the following 

SC&A

l procedures at these uranium foundries is an 

important issue. In 1994, DOE did an assessment and RFP had serious problems in 

We had very sensitive counting systems and urine bioassay at that time, but I am not 

aware of the operational aspects. 

SC&A tdown in

e

ed dose in

RAU: It certainly was true that things were left in disarray. But if there is the need to do a 

e

We note that you do address 238Pu, but you indicate that you do not see the other 

internal contaminants. 

Roger Falk would be the best person to address this, but he is not on the conference

call at this point. 

It is important to note that there were no technical limits unt

: All lung counts, of course, are picking up the Am. If they worked with Am, th

easy to include a dose fo

it was there, we would look for it. But the handling of curium was a small effort at 

RFP. 

S

enough Am in the lung counts?

ORAU: 

year. There was that kind of follow-up for freshly purified Pu. 

: Did you have any individuals who may not have been involved where you picked up 

uptakes of Pu or U?

ORAU: When we went to using germanium detectors and switched from NaI detectors, we 

saw more. 

SC&A: The lack of good radiological contro

their radiation control procedures. 

ORAU: 

: We are not sure how long this went on. In the post-1989 period and after shu

1994, there were lots of materials sitting around that were either leaking or had loos

surface contamination. We are trying to understand how they handled miss

such instances. 

O

dose reconstruction for an unmonitored worker during this post-1989 period, this 

would be calculated using available bioassay data. Things were not just sitting in th
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n

d

bioassay was very sensitive. 

C&A:

8. Wh

reco

information bulletins used? 

RAU: All of the complex-wide bulletins are used as needed to provide general guidance on 

C&A: The response in the Matrix table is OK. 

C&A:

9. The

for

reco r

air c

ORAU: solubility that gives the best outcome to the claimant if the 

solubility is not known, so the TBD no longer needs to give that recommendation. 

NIOSH: 

C&A:

10. The

the t basis is this default value 

recommended when a range of airborne particle sizes has been confirmed at RFP (i.e., 

is approach

recommended to assess intake and dose for systemic organs when using bioassay and/or 

ORAU: 

uation-specific and are 

lines. After the RFP was shut down in 1989, there was a flurry of activity to get

things cleaned up in 1992 and 1993 in anticipation of a restart. It wasn’t until 1993 

that DOE decided they were not going to go back into production. Then radiatio

monitoring programs were established that were based on work permits, fecal an

nasal smears, reentry criteria, and respiratory protection. These were all state-of-theart RadCon procedures. The primary data for any dose reconstruction is the bioassay

data as the highest tier and the

S

ich of the DOE complex-wide technical information bulletins were used in dose 

nstruction for Rocky Flats claims? Under what conditions were the technical 

O

implementation of TBD information and/or to provide a maximizing or minimizing 

estimate if the TBD does not cover a topic of interest. 

S

S

TBD recommends that the dose reconstructor assume that radionuclides are soluble 

the purpose of dose assessment of the systemic organs and tissues. Is this approach 

mmended to assess intake and dose for systemic organs when using bioassay and/o

oncentration data? 

The DR generally uses the

That will be taken out of the TBD. 

SC&A: The response in the Matrix table is OK. 

S

TBD states that particle size and distributions are not available and it recommends

use of the default value of 5 μm AMAD. On wha

don’t measured values take precedence over default values)? Is th

air concentration data? 

Although there is a smattering of particle sizes measured for certain processes and 

time at Rocky Flats, the particles inhaled by a worker are sit
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also that the RFP

particle size data are generally stated in terms of mass median diameter (MMD) and 

r plutonium oxide, the 

AMAD is approximately three times greater than the MMD. Also, use of 5 μm 

rable outcome than using a 

smaller particle size, when assessing doses from urine or lung data (the predominant 

The only situation for which there is a measured particle size associated with an 

e

efault value of 5 μm AMAD

is recommended for all other situations, consistent with the project wide approach. 

Advisory B nt. 

The response by ORAU is OK as long as the DRs base their dose reconstruction on 

bioassay data. The dose reconstruction would not be as good if it was based on air 

SC&A

ORAU imately 92% of claims have some internal dosimetry associated with them. 

This was based on the urine bioassay program until 1989 for systemic organs. DRs in 

in

conjunction with bioassay results. Internal doses below the MDA are covered in the 

We look at as those above the MDA. Some intakes are pre-assigned based on what is 

known about typical doses, like those at the Hanford reactors. 

SC&A

RAU: If you have a Pu uptake, it could still be seen months later, so it might have been 

IOSH:

SC&A: hey monitored short-term workers. We have interviewed a 

millwright who had a Q clearance and was sent into the area were Pu work was 

ongoing. He reported that there was no internal dose assigned that was recorded in 

his personal file. Where would records like this be stored? 

unknown. Therefore, a default value is recommended. Note

not activity median aerodynamic diameter (AMAD). Fo

AMAD results in a neutral or slightly more claimant-favo

situation for RFP workers). 

airborne contamination incident is the 1965 Pu fire, for which the measured particl

size is recommended instead of the default value. The d

oard: Mark Griffin joined the call at this poi

SC&A: 

sampling data. 

: Do you have a sense of the amount of measured vs. estimated data that is used? 

: Approx

Task 5 do not use any of that anyway. Therefore it is not used in the TBD. If they 

have a positive result on their lung count or on their urine bioassay, of course, that 

data is used in the dose reconstruction. Lung counts intakes can be used

missed dose assessment. 

: Big incidents occurred at times when pressure on production was high. As a result, 

people might not have been measured or monitored. 

O

analyzed later and, therefore, not have been missed. If a person was unmonitored 

then OTIB-0002 or OTIB-0018 would be used. 

N It was not unusual to not see some internal dose monitoring. 

We wonder how well t
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SC&A: cords taken away in the FBI raid? 

NIOSH: , we

SC&A: 

ome folks could be missed. 

nter

the area, and they would have been badged. 

C&A:

11. tion

ld be

3 above.” What did NIOSH intend to convey 

in Section 5.2.1.3 which does not exist in the TBD? What value corresponds to X.X μm 

ORAU: been converted to

AMAD. The conversion has been calculated, and the value is 1 μm AMAD. This 

SC&A: e holder for the XX μm value. 

12. In general the values for median MDAs can be considered high, especially for enriched 

the

dos

ndle less-than-MDA

values of data. No special instructions are warranted for this TBD. 

Missed dose is assessed by the DR by censoring the data to 1⁄2 the median MDA 

g 1⁄2

ORAU: There was an effort to consolidate these within the last year.

Were any of these re

ORAU: No, none of the records were taken away in the FBI raid. 

We use all records we can find. If you think there are others and you have copies

certainly would be interested to review them. 

In the process of transition to a more temporary workforce in the post-production 

period, it seems likely that s

ORAU: In the post-production period, only certified radiation workers were permitted to e

S

On page 9 of the TBD in Section 5.2.2.2, it is recommended that the “Dose reconstruc

should assume a 5 μm AMAD, except for fires [for which] a X.X μm AMAD shou

assumed for consistency with Section 5.2.1.

AMAD? 

When this section was written, the value of 0.3 μm MMD had not

value will be inserted in the X.X at the next revision of the TBD section 

It looks like the 1 μm AMAD is going to be the spac

SC&A: 

uranium presented in Table 5.3.1.3.2, on page 14 of the TBD. What guidance is given to

dose reconstructor to deal with results below the MDA when assessing the missed 

e?

ORAU: Task 5 dose reconstructors have general guidance on how to ha

unless the actual MDA is available, in which case the missed dose is assessed usin

of the reported MDA. 
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SC&A

t are below the MDA? Is there a specific 

document for this? 

NIOSH: ent. The last bioassay measurement is used 

and a chronic intake is assumed for the entire period with the minimum being zero 

a triangular distribution. There is not a 

specific document, but there are guidelines to the DRs and we have weekly 

discussions with the DRs to point out new guidelines. Missed dose is calculated by 

running positive readings to determine a chronic dose. We use whichever is larger 

SC&A

C&A:

13. In S s

wer

uran

con

ORAU: ant’s file that the worker was involved in a neptunium

or curium project, the DR may interpret the gross alpha urine in terms of that 

mant-favorable outcome. 

Problem with these questions involved Pu and curium. But until 1963, all results 

applied to U and after that to Pu. But there may be some combined that could be seen 

SC&A

ORAU ied to describe how the data was used at the time. The DR can reinterpret that as they see fit. There is more information to assess than the description 

C&A: Nonetheless, it would be good to clarify this in the next TBD. 

SC&A: 

14. In S rces of interference. The

second interference involves the contribution of the count from 241Am not in the lungs, 

the

contribution may be true for accident or incident situation, but not for the other examples, 

: The MDA is important in developing a missed dose. The MDA for Pu and U was 

very high. How do you deal with results tha

We use 1⁄2 of the MDA to do an assessm

and the maximum at the MDA. We use

(Calculated vs. measured). 

: It would be helpful to see what instructions that you follow. This would help us. 

S

ection 5.3.1.5.1 on page 16, the TBD states that until 1973, gross alpha measurement

e applied for analysis of the urine samples of workers potentially exposed to both 

ium and plutonium. How does the dose reconstructor evaluate potential dose 

tribution from neptunium and curium? 

If there is evidence in a claim

radionuclide if that results in a more clai

by alpha counts and this could lead to missed dose. 

: Would you have looked at the individual case to see if you should look at more than

U before 1963 and Pu after 1963?

: In the TBD, we tr

of the program at RFP in the TBD. 

S

ection 5.3.2.1.1 on page 18, the TBD discusses two sou

but from contributions from contamination on the skin, from material being cleared from

upper respiratory system, or from ingested material. The skin contamination 
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sinc

How does NIOSH deal with this contribution?

All of the interferences, if operative, are claimant favorable when the DR uses the 

recorded lung count data at face value. 

SC&A

SC&A

15. ction 5.3.2.1.2 on page 20 of the TBD, it states that “An underlying 

assumption is that americium remains associated with the plutonium particles in the lungs 

assu

RAU: The basis for the assumption is the observation that americium lung count 

h)

opsy

te that the ratio of measured Pu and Am activities in the 

lungs at autopsy is consistent with that calculated based on the underlying 

SC&A: 

SC&A: 

dep

calculate the activity of U, assuming equilibrium” which was implemented manually 

pers re, based on the

234Th measurement? 

C&A: We would like to get a clarification of who was likely to be exposed to uranium and 

ORAU

you

m. If you have very

freshly purified DU then we also assume equilibrium. 

ORAU: the

nt

e the transit time in the upper compartments of the human alimentary tract is fast. 

ORAU:

: The response in the Matrix table is OK 

:

At the end of Se

until the particles are dissolved or removed from the lungs.” What is the basis for this

mption? 

O

measurements for many RFP workers with confirmed, medium-to-high levels of Pu 

and Am lung depositions remained constant or slightly increased (from Am ingrowt

for over 30 or more years post-intake. In addition, recent USTUR data for aut

cases of RFP workers indica

assumption. 

The response in the Matrix table is OK. 

16. The TBD mentions in Section 5.3.2.2.1 on pages 20 and 21 that “The method to detect 

leted uranium was to detect the 63-keV gamma (doublet) photon of 234

Th and to

238

for special cases in approximately 1978. How did Rocky Flats Plants dosimetry 

onnel infer the lung activity of 238

U, due to depleted uranium exposu

S

who had the potential to be exposed to 234Th?

: During the operation period, the situation was probably super U. The thorium would

come to surface. If you assumed equilibrium and the lung counter saw Th, then

would tend to overstate the 238U if there was super equilibriu

The activity of the DU was calculated from the 234

Th measurement as described in

third paragraph of Section 5.3.2.2.1. When 234Th is in equilibrium with its pare
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section 5.3.3.1 on page 22, the TBD states that “Wound count information is largely 

irrelevant to dose reconstruction.” From other wound decontamination experiences it has 

use

ORAU e

uld

have available wound count and skin contamination data in the claimant’s file. 

238U, the activity of the 234

Th is equal to the activity of the 238U (Section 5.3.2, 2nd

paragraph). 

SC&A: 

17. In

been demonstrated that for some radionuclides, even with decontamination treatment, that

some wound contamination remains in the scars. In case of skin cancer, does NIOSH

wound contamination measurements for dose reconstruction and how is this 

accomplished? 

: The TBD author is not aware of any RFP worker who has had a skin cancer at the sit

of a wound with residual plutonium. If such a case were to happen, the DR wo

Wound count information is largely irrelevant to dose reconstruction. This guidance 

directs the attention of the DR away from the clutter of count data on wound count 

SC&A: rgans.

is

n be

needed for dose

reconstruction for cases of lymphomas. 

ORAU: t does not address lymph nodes 

cancers, but it does state that it should be looked at for specific claimant cases. 

NIOSH:

SC&A: 

18. Fig ounter information, shows that there is no 

data for the right chest. On page 27, it states that “The dose reconstructor should estimate 

data es

NIO

ORAU: for that

period for two detectors and then multiply that calibration factor by 0.43, per the 

reports to the much more relevant information on the wound count reports or 

associated with the wound. 

The reasons and approach in the TBD are OK for the cancers in the systemic o

We agree with the answer, if we are discussing skin cancers. But for lymphomas, it

known that residual contamination, for some radionuclides like Am and Pu, ca

retained in the scars and lymph nodes. A special procedure is

We have a TIB in OCAS that deals with wounds. I

The wound TIB that we are preparing, and that is in OCAS for review, will address 

lymphoma cancers. I can provide copies of this when it is released. 

ure 5C-16 on page 87, dealing with body c

the contribution for the right chest before using the data from the count because the lung

set generally includes contributions from both right and left lungs.” How do

SH calculate the total activity in the lung in this situation?

The DR would refer to Appendix B, equation 5B-6 to get the calibration

following sentence. The net c/m data for the left chest divided by the adjusted 
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calibration factor gives the total americium activity for that count. The DR is directed 

to Appendix B in the third paragraph of Section 5.3.2, for “more detail.” 

SC&A

SC&A

g

cou

“Th If the intake date is different from

the date for the Count Started, the intake date is from the file for a worker with a 

he lung count is used as the intake data.” If 

it was part of the routine monitoring program, why is the date of the lung count used as 

ORAU: 

ricium

the DR was to look at the date to discern whether the computer’s 

calculation was based on a “real” and confirmed intake or just on the count data for 

that date. Indeed, if a routine count gave a “positive” indication, the internal 

dosimetrist would have conducted a follow-up investigation. 

SC&A

for the

fecal

le

SC&A: 

etected

. The more

appropriate method is to use fecal data associated with lung count data. 

NIOSH or ORAU: It is agreed that this section should be enhanced including the 

methods, uncertainties and MDAs. 

: The response in the Matrix table is OK. 

:

19. Were lung counting measurements a part of the routine monitoring program, or was lun

nting performed just in case of an accidental intake? On page 27, it mentions that 

e dose reconstructor should note of the intake date.

confirmed deposition. Otherwise the date of t

the intake date? This does not seem to be the most claimant-favorable approach. 

Lung counting measurements were performed both routinely and for possible 

inhalation incidents. 

The cited paragraph on page 29 pertains to lung count reports generated by the 

ABACOS-PLUS© software, which calculated uranium, plutonium, and ame

activities for every lung count, regardless of the detection of the radionuclide. The 

guidance to

: The response in the Matrix table is OK. One question we have is related to the 

bioassay program. How was fecal analysis done?

ORAU: For the entire operational period, it seems this kind of data was only recorded

accidents. The claimant’s file will likely include it. RFP tried to implement a

program and it was not well received by the workers. Also, it was difficult to get 

consistent results using the chemical analysis they were using. Therefore, they 

backed off of it. There is fecal data for some of the accidents and some data in the

mid-1990s. Missed dose, when assuming constant chronic intake, is more favorab

than looking a bunch of small incidents. 

Fecal bioassay programs need to be enhanced in the TBD in order to provide more 

information about fecal analysis. In cases of incidental intakes of insoluble 

compounds, it is very unlikely that these insoluble forms would have been d

using urine data since the activity expected to be found is very low
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20. On

NIO section? How is NIOSH

planning to deal with the assessment of dose based on co-worker data? Will evaluation of

r concentration data be a part of this effort?

O

SC&A

ORAU d

OTIB

e TBD, but will be

in the new TIB. There is an OTIB-0019 that discusses internal dose. Three TIBs 

ecides if

IB-0038 when it

21. 

ORAU

SC&A: 

page 30, it mentions that Section 5.5.2 is reserved and yet to be determined. What is

SH doing in its revisions to the TBD to expand upon this

ai

RAU: These issues are still to be determined. It is likely that each case, if any arise, will 

need to be researched on a case-specific basis, using whatever data are appropriate 

and available. The next revision may say just that. 

: The response in the Matrix table is okay. 

: Co-worker data is all being analyzed. The work on the wound OTIB got postpone

by work on recent Mallinckrodt Chemical Works reports and research, but the

is pretty far along. The co-worker data will not be addressed in th

have been issued for internal dose and the process recommended takes the data sets 

and does an analysis of them. Values are put in the OTIBs and the DR then d

those intakes are applicable. There is a Rocky Flats internal dose OTIB in 

development. This has not gone to OCAS for review. It will be OT

is released. 

SC&A:

In Attachment A, it is not clear how the volume of urine samples was normalized for 24-

hours excretion. Please explain the two different approaches used for uranium and for 

plutonium and how they were normalized in each situation? 

: In general, routine urine bioassay samples were not normalized to a 24-hour 

excretion, i.e., a standard volume such as 1400 mL. Instead, the worker was 

instructed to collect all his urine excreted over a continuous 24-hour period or two 12-

hour overnight periods, and the worker was expected and assumed to have complied

with the instructions. The median volume of 1350 mL, presented on page 43 and 

obtained from urine data logs for samples submitted 1952–1955 and 1960–1971, 

verifies the validity of that assumption (ICRP-23 Reference Man daily urine volume

= 1400 mL). It was noted (p.40) that volumes less than 1000 mL for Pu routine 

samples were normalized to 1200 mL and, for routine DU samples (p.41), volumes 

less than 1000 mL were set equal to 1000 mL. For EU samples, no such volume

adjustment was made, since only 50 mL of the sample was analyzed. This 

ethod that was used. 

C&A: We still have a question. Will the values in the claimant files be normalized? 

information was presented to describe the m

S
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All values are taken as if they were 24-hour samples. If samples were less, then there 

is a description of what they did in early years to adjust for this. This will not be 

SC&A

22. In Attachment A on page 39, it states that “Depending on the process, spiked samples, 

m

with

plut (first Pu and, later, Pu). All 

plutonium samples were spiked with an internal tracer after 1978...Whether to use the 

ed e depends on the purpose.” This seems to 

suggest that some samples, in the earlier years, may have higher MDA than the ones 

bulated or described with the results of the urinalyses, since the recoveries were 

determined by batch spikes. What is the variability of the recovery? 

ORAU A

SC&A:

SC&A: 

23. In A that

the

ana

dep

whe de (B1) has been assigned in both situations? 

/1/64, coinciding with the wind-down of 

EU operations at RFP, which ended in 1967. Even though it is not likely that one 

to both DU and EU in this period, the DR always has the 

option to assign the analyte that results in the more claimant-favorable outcome, 

when the DR does not have adequate information in the claimant’s file. Note that the 

SC&A: The response in the Matrix table is okay. 

ORAU: 

applicable to more current times. They were usually handled as 24-hours samples, 

but you would use ICRP 23 if you needed to normalize. 

:

sa ples to which a known activity of the analyte was added, were generally processed 

each batch of samples.” On page 52, it states that “Not until 1973 were some

onium samples spiked with an internal tracer 236 242

m ian value of the MDA or the extreme valu

ta

: Indeed, 50% of samples have a sample-specific MDA higher than the median MD

for the process. The table on the bottom of page 44 gives the median and 5th

percentile values of recoveries. The 5th percentile values of recoveries are an 

indicator of the variability at the worst-case end of the distribution. 

The response in the Matrix table is okay. 

ttachment A on page 37 in the Analyte/Method Code table on page 37, it is noted

Method Code B1 appears both for the analyte “Depleted uranium” and also for the 

lyte “Enriched uranium.” Is it possible for a worker to have been exposed to both 

leted and enriched uranium? If so, how does the dose reconstructor distinguish 

ther it’s DU or EU, once the method co

ORAU: The B1 method was used for DU starting 5

worker would be exposed

building number is recorded on the bottom of the Urinalysis Record Card: Building 

44 = DU; Building 81 = EU, which can help the DR to make a correct judgment 

concerning the analyte. 
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TB

ura (and enriched uranium

operations were phased out), the default condition was to credit the result to plutonium.” 

claimant-favorable approach?

The information in Attachment A describes the program as it was conducted in real 

ent approach if needed

unm n

effe ined” 

giv

req he status of obtaining this key information and how are 

unmonitored internal intakes being modeled in the interim?

ORAU: 

or

to assess this situation—use of environmental 

exposures, use of MDA values at the end of employment or risk, and co-worker 

bioassay data. Each case is individual-specific, but unmonitored workers’ internal 

dose is generally assessed through use of maximum and minimum assumptions based 

SC&A: 

POWERP

Slide 12 

SC&A: sumptions

that should be used to calculate the internal dose; for example, missed dose, solubility 

SC&A: 

24. In Attachment A, on page 42 under the section entitled “Gross Alpha (1952–1971),” the 

D states that “The default condition, through 1963, was to credit the result to enriched

nium unless the PHA count indicated otherwise...After 1963

Why does NIOSH feel this is the most

ORAU: 

time at RFP. For any given case, the DR may choose a differ

to achieve a correct compensation decision. 

SC&A: The response in the Matrix table is okay. 

SC&A: 

25. In Section 5.5, Internal Dosimetry for Unmonitored Workers, it is indicated that for 

onitored workers “not involved in an incident,” use of air concentration limits i

ct would be a reasonable approach. However, this section is left “to be determ

en that actual limits applied in the plant are needed, not merely those cited in official

uirements. What is t

Much of that research has been done and will be incorporated in the next revision of 

the TBD. Although it is not clear whether the DR has yet encountered such a case f

RFP, the DR has generic tools

on generic guidelines such as OTIB-2, OTIB-14, and OTIB-18 and, in some cases, 

data from other sites (e.g., for tritium). 

The response in the Matrix table is okay. 

OINT VIEWGRAPHS 

There is limited direction to the dose reconstructor on the process and as

and particle size, and incidental intakes. 
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product. This

may represent no appreciable dose, but it should be addressed... The TBD does not 

ORAU: There is no available information. 

Slide 15 

SC&A: ample data. 

lues of the intake retention fractions (IRF) 

d feces for Type S and high-fired Pu compounds following a single 

inhalation of 239Pu. Urinalysis seems to be inappropriate for the detection of intakes 

here was a really high intake. The fecal 

data are more appropriate for the detection of incidental intakes of insoluble 

compounds, in the first days after the accident. The activity expected to be found in 

feces is about 4 orders of magnitude higher than urine. 

ORAU: e. By

using bioassay, the values will be higher and more claimant favorable. 

NIOSH: In the TBD, fecal analysis results are discussed in a short paragraph. We realize that 

t fecal data is available and how much it was used. 

able.

Slide 13 

SC&A: The TBD needs better guidance for 3

H, and Np, Th, curium and fission

address exposure from the processing of recycled uranium. 

5 μm is acceptable for bioassay, but not for air s

NIOSH The only time that air data is used is for environmental dose reconstruction. For all 

others, we use bioassay or missed dose procedures. 

Slide 16 

SC&A: The TBD assumption for solubility of the compounds is not claimant favorable, 

especially for cancers in the gastrointestinal organs. The graph shows that for 

stomach, small intestine and large intestine (colon) the assumption of Type S 

compound arises with higher doses. So, the solubility should be assigned based on

the type of cancer. 

ORAU: We would deal with this in our guidelines to the DRs and it will be taken out of the 

TBD. We use Type S for colon and digestive tract. We will take this out of the next 

TBD and specific DR instructions will be followed. The solubility should be done 

based on the specific cancer. 

Slide 17 

SC&A: We have provided a table that shows the va

for urine an

of either high-fired Pu or Type S, unless t

We can’t go back and get fecal samples, so we have to work with what we hav

this needs enhancement as to wha

ORAU: Fecal data is rare and difficult to use. Did the fecal sample capture the bolus coming

through the GI tract? If you don’t capture the bolus, then it’s not claimant favor
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We will summarize this, although it might be better to do this in a generic document. 

into what is best with high-fired oxides and we will deal with this in the new OTIB. 

e and feces after 

239

ss

Slide 19: 

SC&A: Slide 19 related to using urinalysis. High uncertainties are 

associate

estimate presents an evaluation of

the extreme conditions for the variables, but does not provide a clear guidance on 

Slide 20 

This has already been discussed. 

Slide 21, 22, and 23 

SC&A: This has already been discussed as well. 

Slide 24 

SC&A: Ingestion pathway of exposure was not considered on the dose assessment; it is 

important especially in cases of cancer in organs of the gastrointestinal tract. 

Slide 25 

SC&A: This has already been discussed. 

SC&A: We would recommend to the TBD authors that they include in the TBD an example 

of a dose reconstruction. This could be placed in the Attachment. This is very 

helpful and should be used in other TBDs. 

NIOSH:

SC&A: Are you using AMAD of 1 μm when dealing with high-fired oxides?

NIOSH: We generally use Type M and 5 μm for most organs other than lung. We are looking

SC&A: It is but a fraction of a 1 Bq intake that is expected to be found in urin

the inhalation of Pu Type S and high-fired compound, and an AMAD of 1 μm 

would be more appropriate. 

NIOSH: There is not going to be a very large difference. It is better to go from bioassay data 

and determine what gets into the organs. The TIB for high-fired oxides will discu

this. We use type M for most organs and Super Type S for lung. 

points out the uncertainties

d with estimates of the high blank, recovery, 24-hour urine volume, and the 

of the median MDA values. Therefore, the TBD

how to apply them. 

NIOSH: The DRs are using a chronic intake with 1⁄2 MDA. This is a much bigger number and

is more conservative. It is not claimant favorable to do individual accident analysis. 

So we feel no action is needed. 

SC&A: 
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SUM UP 

oma and wound TIB to SC&A as soon as it is released. 

compare the entire body of TIBs and will send them to SC&A on a 

SC&

Una rt 

of the call. 

the

workers along with annual exposure summary. The memo stated the following: 

A 0 (zero) is recorded for any of the following conditions: 

2) the exposure was so small it cannot be measured by our current methods 

IOSH/ORAU: The annual report card was a computer generated summary of exposure. If an 

would put a place holder 

possibilities as far as the 

enerated report. It really didn’t mean much. It was just a clarification to 

of all the possibilities for zero.

NIOSH will send the lymph

NIOSH will go through and

CD. 

A will get the online TIBs to Dunstana Melo. 

nswered internal dosimetry questions will go to Roger Falk who missed most of the first pa

SC&A: A memorandum titled Personal Radiation Exposure Summary was issued to

 

 

1) no exposure received 

3) no data is available to make this assessment. 

 

What does no data is available to make this assessment mean?

 

N

individual was not monitored during the year, the computer

zero in the report. The description of a zero covered all the

computer g

the worker
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Date: September 7, 2005 

Time: 1:30

Participants: 

ORAU

NIO H , Greg Macievic

SC&A:

Additio

SC&A

Let’s ju questions and state if the answer on the matrix is satisfactory – then we will 

proceed. If there are any additional questions –then we will discuss these issues as we get to 

eneral Comments:

We

CCUPATIONAL EXTERNAL DOSE (ORAUT-TKBS-0011-6) 

Dosime

RFP Occupational External Dosimetry TBD Conference Call Summary

- 4:00 pm EDT 

: Robert Meyer, Craig Little, Jim Langsted 

S : Brandt Ulsh

John Mauro, Joe Fitzgerald, Ron Buchanan, Tom Bell and Kathy Robertson 

DeMers 

nal Participants: Amy Dean (Scribe), Advisory Board, Mark Griffin, All Wolff 

General Comments:

st go to the

them. 

ORAU G

have used this process previously and this is a good way to proceed. 

O

try Records

SC&A:

was is area of

unc

RAU: The statement, “Further research is necessary to identify exact dates for each 

rmation has been included in the TBD or passed 

on to the Dose Reconstruction group. No further data is available about exactly when 

specific dosimeters were used. 

 

SC& :

2. 7),

regarding skin and penetrating radiation doses, match that given in some of the historical 

1. Page 10 states that further research is necessary to determine which dosimetry method 

used to record a worker’s dose. Has any progress be made in addressing th

ertainty?

O

dosimeter type,” still stands. Dosimeter types were typically phased in over a period

of time (sometimes as much as 2 years). The methods of recording worker doses 

sometimes indicated which dosimeter was used, but sometimes this cannot be 

determined. Where possible this info

A

Why doesn’t information in the TBD, page 11 (and also ORAU-OTIB-0027, pp. 5–
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as:

se) + 

neutron dose 

owever, ORAU-OTIB-0027 (pp. 5–7) uses the open window (OW), cadmium filter 

(CD), and brass filter (BR) to define skin and penetrating doses, but does not relate them

RAU: This will take some additional work to complete. 

C&A:

3. How is it intended that the dose reconstructor correlate the methods in TKBS-0011-6 and 

TIB-0027 in relationship to the recording methods that were in place during the years 

they were recorded, and also changed during the years?

stency of the TBD and the

TIB are OK as written. 

ORAU

the approach agreed to by OCAS to implement 

the external dose TBD. 

SC&A: 

4. How

the total dose in these cases? Note on page 11, Section 6.3.1, that it states that during the 

early period (around 1951–1975) the deep gamma dose and the neutron dose were 

records for the RFP (Owen 1964 and Putzier 1982)? In 1964, Owens defines the RFP 

doses

Skin dose = hard gamma dose + soft gamma dose + x-ray dose + beta dose + 

neutron dose 

Penetrating dose = hard gamma dose + soft gamma dose + 1/3(x-ray do

Hand dose = (hard gamma dose + soft gamma dose + x-ray dose) x 2.5 or 5.0 

depending on area and chemical form. 

H

directly to the quantities as listed above which were in place at the time the doses were 

recorded. 

O

S

O

In regards to questions number 1-3 above, our interest here deals more with the 

consistency of use and approach that is used by the dose reconstructor and how the DR 

reconciles this with relevant historical documents. The consi

: OTIB-27 is a clarification, interpretation, and documentation of selected issues and 

information in Rev 0 and draft Rev 1 TKBS-0011-6. There are no known 

inconsistencies. OTIB-27 represents

In regards to these issues (Questions 1-3), this area needs more attention. It will be 

necessary to determine the dosimetry/records methods used over the years and their 

changes. Page 11 of the present TBD needs further expansion and this information 

will be included in the revised TBD. 

is it intended that the dose reconstructor reconstruct the separate doses to arrive at 
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file

tota

may as

app

RAU: If we need to overestimate the dose in cases where the claimant is approaching the 

e the 

application of all neutrons or all gammas to make it as claimant favorable as possible. 

SC&A: 

whe ic

sys

ORAU: 

6. Has any more information been found regarding the characteristics, performance and 

195 et al.

200

L

neu e

mu o the neutron track glass plates provide similar dose determination 

as would TLDs, or is there a difference?

ORAU: here

the neutron track plates in the LANL TBD for external dosimetry. 

There is a different calibration factor for the track plates and these have been reread 

find it is needed. This information will be

included in the revised TBD. The schedule for the revised TBD is still not certain but 

ssue is operating in concert 

with the SEC petition. Both will come out in December before the board meeting in

corporated into the 

TBD by then. 

recorded as the total penetrating dose, and not recorded as separate doses in the worker’s 

.

Our concern is that if the deep gamma dose and the neutron dose were combined into a

l penetrating dose and that separate doses were not placed in the worker’s file, this 

not be as claimant favorable as calculating the dose that was neutron if that w

licable, and then do a separate calculation for all dose that was gamma. 

O

50% POC, then we will determine the penetrating dose, which will includ

5. Was the Harshaw TLD 600/700 system used for extremity dosimetry during 1983–1991 

n the beta/gamma and neutron dosimetry was being performed using the Panason

tem?

Response is okay 

This is correct, as indicated in Table 6-1. 

SC&A: 

calibration of the LANL processed neutron track “plates” for the RFP from 1951 to 

6? Has the recent Neutron Dose Reconstruction Project Report (NDRP) (Falk

5) assisted dose reconstruction (DR) during this period?

SC&A has reviewed several historical documents looking for information on the LAN

tron track glass dosimeter and there is not much that can be found. Has NIOSH don

ch work on this? D

We have found a lot of new data on these neutron track glass plate dosimeters. T

is information on

and doses changes have been made if we

when it is published, we will incorporate information about these issues that have 

come up and we will address SC&A issues. The TBD rei

January 2006. So it will be important to ensure these issues are in
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NIOSH:

sed to read them back in those days. This process was 

done manually over 2–3 years. They had good quality assurance procedures in place. 

C&A: What sort of error margins did you establish for the overall assessment?

ORAU

nt to contact him on this. 

t the reverse 

70/30. 

ORAU: RFP only recorded

where workers worked. For instance, in Building 771 they were doing wet 

SC&A: tudy, it was noted that in the post-operational era that 

thermal neutrons became a problem. Have you found this to be true in your reviews?

RAU: There are some n/p ratios in the NDRP. For the post-1969 period up to 1970, we 

also back extrapolate to

determine neutron dose for the 1969 to 1970 time period. 

SC&A: 

at bears 

closer looking at. 

NIOSH: 

see a characteristic neutron or

photon pattern. 

C&A: Wasn’t there a criticality control group?

ORAU

C&A: Are you aware of criticality experiments that were done at RFP?

SC&A: The NDRP indicates that these plates have been read. How fragile are the plates and 

is there a problem with physical fading?

NIOSH does not know of any problems with fading. If there is fading in field and 

that same fading is included in the calibration, then there should be no great effect 

and it will likely cancel out any effect. As they re-read the plates they do use 

calibration sets that were u

S

: Roger Falk would be the better person to provide you an answer to this, but he was 

unable to be on the call this afternoon. You may wa

SC&A: Does your data tell you by building number what n/p ratios were used? For instance, 

we have read that the n/p ratio is 30/70 for Pu and for metals it’s jus

We do not have that kind of specific data by building location.

chemistries and in Building 776 and 777 they were working with metal. 

In the DOE 1994 vulnerability s

O

have to be careful how we use the n/p ratios. We can

Have criticality control program accidents been reviewed? Incidents that involve 

alpha/neutron reactions such as perimeter controls on drains are an issue th

If you have a solutions near critical, you would see a lot of higher energy photons 

affecting the film badges. You would then likely

S

: Yes, there was a criticality control group that was assigned for the majority of the 

time. Section 2, page 15, discussed criticality accidents. 

S
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f

looked over all

our data and have n/p ratios from 1970 to current times. The numbers are stable over 

parated.

ill the recorded neutron dose be used in DR that is discussed on page 14, 

Section 6.3.2.2.1, where it states that from July 1984 to October 1984, some neutron dose 

ORAU de that neutron in the total dose. Typically it won’t matter, but we 

still will include it so as to err on the side of claimant favorability. 

C&A: Have you looked at what happened when the badge was contaminated and no dose 

RAU: In such a case we would rely on co-worker or other dose reconstruction processes. 

When such procedures are used, there is a write-up that goes into the individual’s file 

on how the dose was assigned. 

SC&A:

ORAU

SC&A: ews that a number of workers 

shucked the badge and didn’t even use the film badge. Worker interviewees seem to 

ORAU: 

C&A:

the possible manual correction needed and 

,

corrections. If such manual corrections 

 

NIOSH: Yes, we are aware of these experiments and they are covered in the TBD. In terms o

the NDRP, quality control is covered in the appendices, as well as how they did their 

statistical analysis. NDRP has n/p rations for different buildings. We are working on

OTIB-50 that will go over the ways DRs will use NDRP. We have

the years. In more recent times, the n/p ratios have been conjoined, but in the earlier 

days they were se

SC&A: 

7. W

was recorded, but the gamma and total dose was zero?

: Yes, we will inclu

S

was found in the records associated with such contamination?

O

In the DOE dosimetry file do you have both measured and calculated dose?

: Yes, the file would include any pertinent dose data... 

We have heard recently during our site expert intervi

indicate that this could be a problem that will create gaps in their records. 

In such cases, if they are identified by the claimant, we use n/p ratios and/or back

extrapolate to get a dose. We also look at claimant interview and then we compare 

what they are telling us with the dosimetry file. 

S

8. How are the dose reconstructors handling 

discussed on page 14, Section 6.3.2.2.2, where it describes a possible manual correction

but ends with the statement that dose components were not provided in the letter and, 

therefore, were not made?

It is unclear what is meant by possible manual

were done were they placed in the person’s record?
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g

1984

C&A: Do you have any other information on why that occurred or what was supposed to be 

put into the file? 

OR was.

SC&A:

9. e

 This is okay – no further questions. 

ORAU

r the dose reconstructor. 

ation workers included on page 18, Section 6.3.4.2, 

where it states that the NDRP (Falk et al. 2005) will cover 1951–1970 when NTA film

wor

crit sources, etc.] besides PU at the RFP during this time 

period when NTA film results/procedures are in doubt?

ORAU: 

e available films and reassess the dose for 

those individuals only. The scope did not include other workers, unfortunately. 

SC&A: 

11. Can study by 

University of Colorado-HSC mentioned on page 18, Section 6.3.4.3, that appears to be 

helpful in DR? 

Have you received the job exposure matrix? 

ORAU: The deep dose would be used and applied to compute the most favorable dose. The

dose would be used unless there was information to the contrary in the file. Applyin

the dose in 1976 would be claimant favorable if the dose was really incurred in

through 1986 and appropriate, given lack of contrary information. 

S

AU: No, we just found these files, but they do not provide what the correction factor

Why isn’t more information provided besides the exchange history (i.e., quarterly dos

history, etc.) that is discussed on page 18, third line list: “1977-present, dosimeter 

exchange history?” 

: The wording in the text is unclear. What is intended is that for 1977 onward, the 

dosimeter-by-dosimeter record is available fo

SC&A: 

10. Why aren’t uranium and non-radi

was used for neutron dosimetry? How is NIOSH handling dose reconstruction for 

kers exposed to other neutron sources [(alpha, neutron) reactions such as in UF4, 

icality experiments, calibration

Uranium workers and non-radiation workers were not monitored with neutron film. 

The scope of the NDRP is to reevaluate th

Early neutron film response correction (Table 6-18) does include a range of neutron

spectra (metal, waste, salts, and PuF4). In the later years neutron dosimetry systems 

were calibrated with a variety of neutron spectra. 

NIOSH make available to SC&A the DOE-funded Job Exposure Matrix
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We do have copy of the report that Dr. Ruttenber prepared. We are trying to get a 

copy of the job exposure matrix, but this is currently in the hands of our lawyers. It 

er.

12. How

ind

calibrated to) on temporary or overtime assignments, which the dosimetry department 

could not detect? Does NIOSH consider this an occurrence that happened often enough 

significant underestimate of dose to workers because of the change in the 

dosimetry response? Does this require some modification to the NIOSH decision stated 

n page 27, Section 6.6.3.4, that assumes that each worker stayed in the work area where 

the badge was calibrated and therefore no corrections for photon energy are needed?

SC&A: 

NIOSH:

SC&A: ant

in

ork in other buildings on a temporary or overtime basis is likely to 

ORAU: 

might have pertinent information and we are committed to go back and make 

changes, and will provide SC&A a copy when we get it. Dr. Ruttenber did some

work in the mid-90s, but NIOSH does not have this eith

SC&A: 

has NIOSH dealt with the situation discussed on page 19, Section 6.4.3, that 

ividuals sometimes worked in other facilities (beside the one their badge was 

to create a

o

During site expert interviews there have been discussions of people working part-time in

another work area or working after-hour duties so that their badge might not be properly

calibrated for these non-routine work locations. 

NIOSH: It seems unlikely this could have any significant effect on total dose. 

Would records help on this?

We tend to work more with general work assignment rather than day by day changes.

We understand that part of what Dr. Ruttenber did may have included such data and

that is why we are interested. 

We are hearing that there is about a 50/50 split. Some workers work all over the pl

and others work more in stationary locations. For instance, machinists often stayed

one area due to special training needed. 

ORAU: We have found that in general, lower-level workers stayed put, but the upper-level 

workers moved around. Everyone had a badge. Therefore this is not a problem in

determining dose. Most workers spent their time in areas in which their badging was 

appropriate. W

result in dose which is a fraction of the dose received from the routine work 

assignment. Given the number of claimant-favorable assumptions in the dosimetry

records, it seems unlikely that large underestimations of dose occurred. 
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C&A:

13. Wh

use

wit

ORAU: as zero. Zero is a potential missed dose 

and NIOSH would assign dose based on the MDA at the time. For missed dose we 

ommon Issues

S

y isn’t the 10 mrem minimum detectable dose discussed on page 20, Section 6.4.4,

d as the dose of record rather than reporting this as a zero dose? Isn’t this inconsistent 

h the other AEC site TBDs and not claimant favorable?

It was RFP’s policy was to record 10 mrem

assign the LOD/2 and a geometric SD of 1.52. The dose of record is the dose 

reported in the worker’s dosimetry file, not the TBD. The TBD reports the practices

of the time. 

C

C&A:

14.

 Why would there not be a date associated 

RAU ate" is the nominal date on which the badge wear-period ended. This does 

ation date. The

 to the file may not have had a date associated with it 

etimes just a year). Hire dates were sometimes corrected because of prior 

service and termination dates corrected to allow use of remaining vacation or sick 

leave. These and other exceptions (which are not well understood) result in date 

discrepancies in many of the dosimetry records. 

SC&A: ssigned when they got their dose. It appears that 

it could vary from 30 to 90 days. The beginning date could go back a number of 

For example, if the activity date was something like August 31st, the badge might be 

picked up on August 29th and processed a week or so later. The next badge would 

le of

S

Page 21, Section 6.5.2, second paragraph states that “If no activity date is associated with 

a dose record...” What is an “activity date”?

with a dose record? Also, page 15, Section 6.3.3.1.3, mentions that the Activity Date 

may be outside the employment period. What does this mean?

O : "Activity D

not account for weekend days or holidays, which delayed the badge pickup. This is 

discussed (somewhat vaguely) in Section 6.3.3.1. Later when "start date" and "end

date" were incorporated into the new database, it was necessary to synthesize these 

dates from the activity date and the exchange period. Thus, it was possible to have a

"start date" before the hire date or an "end date" after the termin

addition of "correction" record

(som

There is a concern that dose is not a

years. 

ORAU: 

pick up for the remaining days of August. 

If a person started part way through the quarter, they would only have a partial dose 

for the quarter and they might have to continue to wear their badge into the midd

the next month. 
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e

dose is included even if it’s outside the date of employment. If a previous dose that is 

have

e

. However, if pops up in 1986, they would then apply it in 1976 

when it was reported. This would be most claimant favorable. 

SC&A: 

the DR process discussed on page 23 regarding exposure geometry, why does NIOSH 

consider the assignment of ISO or ROT instead of AP geometry as always most claimant 

0%

NIOSH ometry effect of using frontal A-P and it is not always 

favorable. Tim Taulbee is working on a draft TIB that will address this issue. This is 

B 

Photon Do

There are situations where RFP recorded previous dose and give it a date prior to hir

date so that they could see they came from somewhere else. Typically we put in the 

earliest date it was recorded and this could be different from the actual date. That 

not acquired at the RFP is found, it’s unclear how this would be handled. If it is in 

the RFP employment period, it would be used. It depends on the specific case and 

scrutinized and included in some fashion. 

There was one working activity date assigned in 1986, but they really could

been received as early as 1976. It is best if the dose they received is recorded in th

period it was received

15. In

favorable (i.e., for anterior-seated cancers)? 

It is realized that some claimants could be exposed from back when their cancers might

be due to higher front-sided orientation. So their dose would be greater if it were a 10

A-P exposure. 

: We have looked into the ge

a hot issue and will be addressed generically. We are working on finalizing the TI

soon. The addition will also go into the IG for use by all DRs.

se

page 24, 6th line up from the bottom, shouldn’t it read ...the 30-250 keV photon..., 

ead of ...the 30-50 keV photon....?

SC&A: 

16. On

inst

RAU: Thisis a typographical error that will be corrected. 

SC&A: 

17. Has NIOSH considered the need to take into account during DR the effect of shielding 

bec

that the photon exposure spectra, assumed to be mostly in the 30–250 keV range, is 

claimant favorable because of the higher REF for this energy range. 

O 

where photons with energies >250 keV may contribute more to the final effective dose 

ause of their greater penetrating power? The TBD states on page 24, Section 6.6.1.1, 
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The REF reference came from a HPS Journal article of July 2005, pages 3–32, that 

looks at the radiological effectiveness factor for 30 keV photon and higher. Would it 

ORAU

laimant favorable. There is about a 15%–20% difference where the 250 

keV peak occurs. Generally this is not large problem and the 30–250 keV is most 

ovebox material shielding was 

considered. Uranium was typically processed outside of gloveboxes and was not 

subject to this ef

 

It is not clear where the referenced text, "...higher REF for this energy range," is 

located. The most claimant-favorable photon-energy range is used in DR. 

C&A:

ation on how angular dependence has been 

e Panasonic system response decreased as the angle increased. 

 18, states that the recorded dose of record is 

ng

(4) TLD beta/gamma badges?

The TBD's position on angular dependence is stated in Section 6.6.4.1, “There is 

Panasonic dosimeter, increase the dose only slightly, not including this correction 

e knows for sure.

We would like to recommend that NIOSH develop some internal guidelines on 

angular dependence and the impact that has on gamma, beta, and NTA neutron 

IOSH: neric problem that we intend to handle on an overall basis. NIOSH will 

SC&A:

be more claimant favorable to assign the higher of the two?

: Photon dose correction factor is used for 30–250 keV spectra in most cases. This is

considered c

claimant favorable. The energy distribution for plutonium (shown in Table 6-6) does 

include the effect of spectrum hardening. Gl

fect. 

S

18. Could NIOSH provide more inform

addressed? On page 27 it states that the dosimetry response increases with angle, but on

page 28 it states that th

The Y-12 ORAUT-TKBS-0014-6, page

likely too low at non A-P angles for beta/photon doses. What is NIOSH’s position on 

a ular dependence for: (1) NTA film, (2) beta/photon film, (3) TLD neutron, and 

ORAU:

insufficient data to identify an angular dependence correction to apply to any of the 

dosimeters. Because any correction would reduce the dose or, in the case of the 

factor is generally claimant favorable.” The statement on page 27 (response increases

with angle) is for the film badge, while the statement on page 28 is for the Panasonic

badge. This data is relevant to the RFP badges, while the Y-12 Plant data is relevant

to their badge. The TBD's analysis of angular dependence is presented in Sections 

6.6.4.1, 6.7.4.1, and 6.8.4.1. With little data on the effect of angular dependence, 

especially in the early days, probably no on

SC&A:

dosimetry response. This could be an important factor. 

N This is a ge

check on this and let SC&A know about the time line for how and when we will 

address this.
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SC&A: 

19. Why is the uncertainty for low doses of 1 and 2 mrem less than for the higher doses of 

mre 29, show the opposite

characteristics. 

ORAU: 

Dose (mrem)

photons (i.e., 1 x1.00 = 1 and 2 x 1.00 = 2, but 10 x 1.23 = 12 and 100 x 1.23 = 123 

m, etc.) in Table 6-12, page 30? Tables 6-10 and 6-11, page

This is a result of rounding the result to the nearest mrem value (consistent with RFP

dosimetry data). The actual multipliers are: 

 1983–1989 1990–1998 1999-present

1 1.27 1.27 1.23 

2 1.24 1.24 1.19 

SC&A:

pro

ORAU: 

comp case with only summary data. 

C&A:

21.

ORAU:

5 1.23 1.23 1.18 

20. Please explain how Table 6-3 will be used by the dose reconstructor to amplify what is 

vided on page 21?

Table 6-3 can be used in determining the maximum number of zeros to assign a non-

 

S

In calculating the exposure geometry factors, what dose rate and exposure durations were

assumed? From what source was this data derived?

 Since the results are presented in fraction of the dose received via each geometry, the 

dose rate is irrelevant. Likewise, the hands-on work time is presented as fraction of 

the total work time; thus, the actual exposure time is irrelevant. What is important 

here is the relative ratio between the hands-on dose rate (AP geometry) and the other 

6

ORAU ustrial radiography appears to have been

performed by outside contractors. RFP-owned sources may have been possible, but 

nt.

dose rate (ISO or ROT). This ratio (AP/ISO or ROT) is 4 for the line source and 1

for the plane and point source geometries.

SC&A: 

22. Did RFP ever acquire any sealed sources for use in industrial radiography such as 226

Ra, 

137Cs, or 90Sr?

: From the information we have found, ind

we have no such documentation. As far as we are aware of there was an NDT pla
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vault

NIOSH:

Neutron D

There were very large MeV x-ray units used to x-ray pits, but this was done in a

where no one entered. 

SC&A: Are you aware of how many of these MeV x-ray units there were and what their 

milliamp output was? Is there any documentation on this? 

No summary documents have been found, but we can look at this in more detail as 

well as look in the incident records. This has not been done yet. As far as neutron 

generator(s), we know it existed but we don’t know applications. 

ose

SC&A: 

23. Why is there a difference in the limit of detection (LOD) between page 33, Table 6-16, 

ts the limit of detection (LOD) for neutron NTA film for 1961 as 96 mrem and 

Table 6-17 on page 34, where the LOD is listed as 120 mrem?

RAU: Table 6-16 reports the values determined by the NDRP. The paragraph immediately 

e

d

C&A:

24. Page 34, Table 6-18 lists the potential missed neutron dose below 800 keV for early NTA 

claimantfav e

of 6 selected (i.e.,

1/1-0.60) = x 2.50)? In addition, even if the value of 56% was selected for the percent of 

eutron dose missed, why wouldn’t the multiplying factor be: 1/(1-0.56) = x 2.27 instead 

of 1.79?

ORAU

SC&A

information will be used until the NDRP report 

is finished. The NDRP report (Falk et al. 2005) was available as of February 7, 2005. Is 

resu

This question ties together with Question #10. The NDRP addressed mainly the Pu 

workers. Therefore, it is not helpful for non-Pu workers and U workers. Is there 

which lis

O

following the table sites a study performed at that time. The more claimant-favorabl

value was selected. Table 6-17 summarizes values that should be used by the DR an

takes into account this more claimant-favorable selection. 

S

film to range from 16% to 60%. However, the text below it selects 56% as the

orable value with a resulting multiplying factor of 1.79. Why wasn’t the higher valu

0% missed neutron dose with a resulting multiplying factor of 2.5

n

: This is an error. The commenter is correct that the maximum value in the table is 

60% and the multiplier should be 2.5. This will be corrected.

:

25. Page 35, Section 6.7.3.4, states that other

the NDRP currently being used in DR? Has it resulted in any significant changes in the

lts of past or current dose calculations? 
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nything in the works to extend this to other areas where there might be significant 

changes in results of past or current dose calculations? 

NIOSH

ed cases since it was issued, and final 

implementation issues are being addressed. Past cases were worked with highly 

mber

l

id take estimates at Fernald and apply 

those estimates to workers. We used 10% of gamma dose and applied that. We do 

need to look at EU to estimate neutrons there which is an analogous process. Advice 

to the DRs will be given as our efforts prog 

Electro

a

: DOE did not release the NDRP data to NIOSH until sometime in June 2005. The 

NDRP data has been used for select

minimizing (for compensable cases) or maximizing (for non-compensable cases), and 

thus are not expected to have a major impact on past dose reconstructions. A nu

of cases that could not be worked with maximizing or minimizing assumptions will 

now be able to be worked as a result of the NDRP data. DOE has not and likely wil

not fund such additional studies. On DU, we d

ress.

n Dose

:

Page 37, next to last sentence states that the RFP had problems with elevated beta dose 

rates from contamination on leather gloves worn during foundry operations. Did these 

workers routinely wear wrist badges

SC&A

26.

so that these doses could be accounted for? If not, 

Thi

ORAU: 

SC&A: 

27. Pag tes that the assembled 

badge displays severe angular dependence to beta exposure and that for TLD badges it 

% to 59% of the true beta dose at +

how will DR be performed? 

s is related to extremity dose as well. 

It is unlikely that wrist dosimeters would measure the dose resulting from

contaminated gloves. This issue must be addressed for those relatively few claims

resulting from extremity or skin cancers originating on the hands.

e 40, Section 6.8.4.1, addresses angular dependence and sta

might record only 36 30 degrees. Wouldn’t this need 

to be addressed, especially in the extremity section where the worker’s normal 

ovements would not tend to average it out? 

ORAU:

(DOE 2001) it was determined to be inappropriate to use a correction factor 

for body dosimeters. 

m

This is related to extremity dose as well. 

 It is relevant to address this issue with extremity dosimetry. Based on DOE Good

Practices
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SC&A: 

28. Page 41 Ta 1.12.

However, t r and do not 

match Colu

ORAU: This is 1.1215,

and 1.1 w the

multipl ultipliers

are used in the DR process, they are the result of interest. It is inappropriate to use 

more than three significant figures, however. 

SC&A: 

29. How can V f

Section 6.8 rom

contamination. However, Table 6-26 on page 43 shows that the VARSKIN Mod 2 only 

includes one ( Th at 3.40E-07 Ci/gm of DU) of the two most active beta isotopes [the 

other b

DU is not i

is dealing etic 

contributo

ORAU: The co

skin do d

version t code

to estim

include

input p

VARS onstructors who process RFP cases will be made aware 

of this issue through Task 5 meetings, and the next revision of ORAUT-OTIB-0017 

will inc

be incl

 

Unmonitored Ind

, ble 6-24 footnotes state to multiply any dose greater than 2 mrem by

he fourth column contains numbers that do not use this multiplie

mns 2 and 3. Why is this?

a function of rounding. At 2000 mrem the factors are actually 1.1215,

156. These sections should probably be rewritten to correctly sho

iers as reasonable approximations based on these tables. Since the m

ARSKIN be used if it leaves out 234mPa? Page 42, the last paragraph o

.6 recommends using the VARSKIN software to calculate skin dose f

234

eing 234mPa at 3.40E-07 Ci/gm of DU (the third isotope, 238U at 3.40E-07 Ci/gm of

tself an active beta emitter)]. Page 37 states “Thus, for depleted uranium, one 

essentially with 2.29-MeV (Emax) beta particle from 234mPa, the most energ

r to the beta exposure.” 

mmenter brings up a very good point. It is necessary to include 234mPa in any

se calculation that is performed. It will be necessary to implement an update

of VARSKIN, modify the code to include this isotope, or use a differen

ate skin dose from skin contamination. The code package VARSKIN 2 

s VARSKIN Mod 2 and SADDE Mod 2. The latter code allows the user to 

ertinent information for beta-emitting radionuclides that are not in the 

KIN library. Dose rec

lude a reminder to dose reconstructors that all significant radionuclides must 

uded in dose calculations from skin contamination.

ividuals

SC&A

at

sho y

195

mo

:

30. What other measures, such as the NDRP, are being implemented to identify workers th

uld have been monitored, but were not? Page 42, Section 6.9.1, states that in the earl

0s, the only groups expected to receive doses greater than 10% of the RPG were 

nitored.
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For some plutonium workers, neutron monitoring was not provided 

gnificant neutron fields from the 

l

P document entitled “Film 

Procedure Timeline” for the Neutron Dose Reconstruction Project 2003 

onth.

of 1362 workers were badged throughout the year

and approximately 300 NTA films were processed per month (therefore, 300 

rocedures in place 

during this period at the RFP were adequate to ensure that the workers who 

ion doses (especially neutron doses) 

ere actually, and adequately, badged. It would appear that the “10% of the 

ers

that

ORAU: The sta rs not

expected to receive more that 10% of the occupational limit are not required to be 

ss

0

ion 6.9.1 or other appropriate dose, such as the 

maximum dose for later years. If later years/work history/dose history indicated that 

a. The NDRP report (Falk et al. 2005), page 1, states that: 

until the early 1960s, and their dose of record may not include 

significant contributions from neutron exposure received prior to 

being issued a neutron dosimeter. These workers included most of the 

employees working in Building 71 (now Building 771). Only a small 

number (10–18) of these employees were monitored for neutron 

exposure, and that monitoring was only during the period October 

1956 to September 1957.” 

Operations in Building 71 involved chemical processing of plutonium

in acid solutions and resulted in si

alpha-neutron reaction with light elements, especially from the 

plutonium tetrafluoride compound. No evidence has been found that 

neutron shielding was present for these operations until the early to 

mid 1960s. 

b. Column 2 of Table 2, pages 17 and 18, of the RFP annual radiation exposure

report for the year 1984 (Radiation 1984) provides the total badged personne

per year for the years 1953–1984. Another RF

(Rocky 2003, pg. 6), provides the number of NTA films processed per m

Analyzing the two tables for the years 1959–1969 show that out of the total 

workers badged, on the average only approximately 25% were issued NTA 

film badges (over the 10-year period, it ranged from 9% to 50%). For 

example, in 1960, a total

workers were badged for neutrons throughout that year); this results in a ratio 

of 300/1362 = 22%. 

 

These facts raise the question as to whether the policy and p

were at risk of receiving significant radiat

w

RPG” policy did not necessarily provide adequate monitoring for all work

needed it.

tement as quoted is correct, and applies today for many sites─worke

monitored. The co-worker dose estimation process is under development to addre

this issue. If the EE job description and/or work or dose history in later years 

indicated potential dose during this time, the EE would be assigned either the 60

mrem indicated in the TBD Sect
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y

ance in section 6.9.2.

,

t. For the early days 

ant-favorable doses based on location, job, and gamma 

t are not in the NDRP but were in neutron buildings 

of neutron dose is based on job description, NDRP data, building, and 

priate. Neutron dose assignments are made based on multiple lines of

evaluation and are thought to be claimant favorable. 

SC&

issed dose if they 

o

NIOSH: Ou onal

ext ill use maximum dose for later years 

 the earlier years. Or we will assign the 

By 1953, 1/3 of the workers were badged 

Dr.

wo

ND

det ich value is used.

no dose was likely and the EE was not in a production area, then the EE is routinel

assigned ambient dose per the guid

The NDRP has addressed neutron dosimetry gaps in the application of their protocol

and the co-worker process will provide some additional inpu

with limited dosimetry, it will be necessary to apply a co-worker process. Prior to the 

NDRP data, claimant-favorable estimates of neutron dose (for non-compensable 

cases) were used. The NDRP data not only included the measured neutron dose, if 

any, but also estimated claim

dose. For those few cases tha

with jobs that may have involved exposure, the neutron to photon ratios or other 

method is used to estimate neutron dose. 

It is not clear from the numbers provided in this comment that the conclusion ("...did

not necessarily provide adequate monitoring...") follows. This is, however, the 

history that we are faced with. It will be necessary for the co-worker process to

provide claimant-favorable dose reconstructions for unmonitored individuals. 

Application

era, as appro

A: There is a concern that some workers may not have been adequately badged or 

worked in areas where later it was found that there really was some dose and they had 

not been badged. Also we are interested to know if you assigned m

lost badge. What is your comment about people in buildings where it was originally 

th ught that there was no dose?

r Task 5 efforts have been trying to address this, and co-worker data and additi

ernal data are just coming online. We often w

if it looks like they were doing this work in

administrative limit in cases where there is no other data. 

At Y-12 - Only 13%–15% were badged. 

Based on the claimant populations at RFP (not plant as whole) – external 

gamma started in 1952 

By 1964, +95% were badged with a combined film/security badge. 

Ruttenber might have this information. Dr. Ruttenber concentrated on those 

rkers with a 1 rem total dose, however. So he might not have all the data. The 

RP gives some reconstructed gamma dose – wrist to WB – that will help to 

ermine wh
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val

SC&A: If t

the

IOSH: If we don’t have it, we hold up doing the claimant’s dose reconstruction until more 

validated.

SC&A: 

NIOSH:

us

resentation to show

when they apply n/p ratios (1979–1976). OTIB-50 will have this information, and 

Extremity

Co-worker study – we are at a state where we have 52 thru 78 and now it is being 

idated. This is coming on line. 

here were situations where there is no co-worker data, but later it was found that 

re were radiation fields – how will that be handled? 

N

data comes in. It’s hard to comment until we see that the co-worker data is

Who is generating the data?

This is being developed as a joint effort between Task 5 and Task 3 (site profile 

folks). In Area 38 n/p ratio is used to estimate dose. The present TBD does not foc

on n/p ratio, however, and more work will be done. For the most part when we 

cannot use the NDRP data, they will call out individual timeline p

this can also be summarized in TBD revision in December 2005. 

Dosimetry

IOSH currently working on TBD changes to make more clear how the following 

es are addressed in, Section 6.10, page 43? 

SC&A: 

31. Is N

issu

a or

this only for beta/photon? Was any NTA film used in 

b

e whole-body (WB) dose 

to the hand if he/she was working on a task requiring hands-on work? [The 

NDRP (Falk et al. 2005), page 22, list the average WB photon dose response as 

only 40% of the wrist photon dose. Additionally, in most cases the dose to the 

hands would be greater that the dose to the wrist (Mann 1964).] 

c. Valid hand-to-wrist ratios - It states in this section that the details on the handto-wrist ratios are not available. In a letter by J.R. Mann (1964), it is stated that 

the hand dose is 2.5x or 5x the wrist dose, depending on the work location. How 

can it be determined if these ratios represent the working environment at RFP if 

we do not know their values (none were presented in the TBD) or how they were 

derived? 

. NTA film in wrist badge - Between 1951 and 1970, film dosimetry was used f

extremity dosimetry. Was

the wrist dosimeters? If not, how will the extremity dose from neutrons (which 

could have been greater than the whole-body dose for those workers handing 

neutron emitting material) be reconstructed?

. Hand-to-wrist-to-WB ratios – If a worker did not wear a wrist dosimeter, how 

can it be considered claimant favorable to assign only th
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on states that additional

s is required before shallow doses to the 

extremities can be reconstructed. The last paragraph on page 37 states that the 

beta dose to the hands could have been a problem because of considerable handson work performed at the RFP. Has any additional information been acquired and 

RAU: It is not clear how many claimant cases will require careful extremity dose

e

ey

ny further. If neutron film was used in extremity dosimeters, then 

these results will be evaluated. If not, it will be necessary to develop and apply 

neutron to gamma ratios appropriate for extremity dose at RFP. 

ity dose. If hand-to-wrist measurement studies conducted at RFP cannot be 

We

P

omment. To 

ion has been acquired relative to extremity dosimetry. 

y that there will be hands on work at RFP so extremity dose 

?

em down to put in the 

d. Additional information - The last sentence in this secti

information on the hand-to-wrist ratio

is it being used?

O

reconstruction. It may be possible to use maximizing and minimizing assumptions to

address the appropriate cases. If additional information is required, this will be 

developed and documented. Detailed analysis of extremity dose is a concern. W

will do a maximum/minimum and if no one falls into the middle category, then th

don’t pursue it a

The assignment of body dose to the wrist for those not monitored with extremity 

dosimeters was the policy at RFP. We cannot change this fact; only develop dose 

reconstruction bias correction factors appropriate to provide the best estimation of the 

extrem

located, it will be necessary to use studies performed elsewhere in the industry.

must assess the appropriateness of these studies to the exposure conditions at RF

and consider the data that is included in the documentation cited in this c

date, no additional informat

SC&A: But isn’t it more likel

could be more of a problem

NIOSH: Hand-to-wrist ratios have been seen but we have not distilled th

TBD. This could be done. The body-to-wrist ratio would come from claimant data 

we have to date. If the worker did not wear wrist badges and had a body skin dose, 

we would have to review and evaluate that closely. But we do not assign whole-body

dose as the wrist dose if the wrist is expected to be higher. 
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ENTARY EXTERNAL DOSE INFORMATION FOR 

OCKY FLATS PLANT (ORAUT-OTIB-0027) 

SC&A: 

1. Page 5, first paragraph, third line: shouldn’t this read Rocky Flats Plant – Occupational 

Yes, this appears to be a typographical error. It was a typo but should not have been 

the Environmental Dosimetry TBD; it should have been Rev. 01-A, March 3, 2004, 

blished changes in Rev. 1 of the

SC&A: 

Pa R)

an

deleted?

 ORAU: Yes, this appears to be a typographical error. Error has been noted and has not 

PO

We have covered m

Slide 3

RAU: We do look at the potential external dose from RU, but we don’t consider it to be an 

internal problem. The potential for possible external dose is not covered in the TBD. 

The only case where RU would be involved is in operations with any significant 

concentration of RU. This would be possible for operations, such as vacuum melting, 

and then this will be addressed. 

C&A: In the Ruttenber 2003 doc, there was a discussion about a large number of neutron 

and gamma doses in 1976 that were erroneously entered. Do you know anything 

about this?

ORAU: Al – when the switch-over occurred, there may have been some who did not work in 

1976 but got a dose. But we use it as if it were real. This has only happened in a 

couple of cases and the incidence is very small. It may be more a case of effective 

dose rather than a missing dose. 

SUPPLEM

R

Environmental Dosimetry (Rev 01).

2

? 

ORAU: 

of the Technical Basis Document for Rocky Flats Plant-Occupational External 

Dosimetry. The purpose was to document unpu

External TBD. 

2. ge 5, in the pre-1960 section lists that skin dose was determined by (OW + CD + B

d also by (OW + CD). Isn’t the BR term in error in this paragraph and shouldn’t it be 

affected dose reconstruction. BR will be deleted in next revision.

WERPOINT SLIDES REVIEW 

ost of these 

Issue: Potential external doses from RU not considered. 

O

S
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Slide

UM UP

(1) Related to Q1 – Further research will be done on date-of-use of each dosimeter system. 

)

) ces,

(8) TLDs worn under the lead apron - This has not been addressed in the TBD. Pantex 

ith Iowa as data coming out of Pantex. 

7

SC&A: The NDRP report shows that the later neutron threshold was 0.48 to 0.75 keV. How 

will this affect the neutron dose (got these from historical literature – Mann and Boss

in 1963)? However, the early threshold was from 0.25 and 0.4 keV. 

NIOSH: I think they used a threshold of 800 keV and did not use the lower ones. 

S

This information will be put in the revised TBD. 

(2) Questions 1, 2, and 5 will be reviewed. 

(3) Related to Q6 - More information on neutron track plates from LANL will be included 

in the revised TBD. 

(4) Related to Q18 - Angular dependence – General guidelines on this issue are being 

worked on and they will let SC&A know when the OTIB or other studies will be done. 

(5) Related to Slide #3 – We will look at Y-12 for EU data and use in RFP TBD if 

appropriate. 

(6 Related to Slide #3 – More information on RU and processes that concentrated it will

go into the revised TBD 

(7 We will include any information found on industrial x-ray units, radiographic sour

and neutron generator in the revised TBD, and send to SCA. 

data supported the wearing of TLDs under the lead aprons. Determine if it is 

applicable to RFP? It came up w
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s reviewed to date are described below. Site profiles completed to date by 

e SC&A team include Bethlehem Steel, Mallinckrodt Chemical Works (MCW), Iowa Army

L).

RAUT-OTIB-0006, Technical Information Bulletin: Dose Reconstruction from 

Doses in Diagnostic X-ray Procedures

incoln and Gupton 1958). Lumbar spine x-ray dose has been integrated into the latest revision 

x-

• Annual photofluorography examines were applied to workers at RFP from 1953–1968 

achine records for equipment were not readily available prior to 2001. As a 

result, the RFP site profile assumed the default assumptions outlined in ORAUT-OTIBhe assumption was also made that technique factors 

at RFP were similar to those at other DOE sites (Furman and Lopez 2004, pg. 9). 

r-anterior

.

 

Lu e r

and lat ployment medical examination. Ultimately, 

occupational medical x-ray dose should be determined based on the individual’s medical record. 

ATTACHMENT 5: CONSISTENCY BETWEEN SITE PROFILES 

The default site profile assumptions and methodologies were reviewed for the Rocky Flat Plant 

and other site profile

th

Ammunition Plant (IAAP). Hanford, Savannah River Site (SRS), the Y-12 National Security 

Complex (Y-12 Plant), and Idaho National Engineering and Environmental Laboratory (INEE

An additional site profile in the process of review is the Nevada Test Site (NTS). 

Occupational Medical Exposure 

The basis for occupational medical exposure for the RFP site profile and other site profiles is 

O

Occupationally Related Diagnostic X-ray Procedures. In the absence of site specific x-ray data,

the RFP site profile relied on values provided in this document for chest x-rays. For guidance on 

lumbar spine exams, the RFP relied on Radiation

(L

of ORAUT-OTIB-0006 (Kathren 2005). This information should be considered in subsequent 

revisions of the RFP TBD. The differences identified between the RFP site profile and other site 

profiles include the following. 

• The RFP site profile did not include lateral chest x-rays. Since there were not lateral

rays considered, there was also no correction for lateral chest x-rays. 

• The Rocky Flats TBD assumed an anterior-posterior and lateral lumber spine x-ray 

from1952–1974 upon first hire (Furman and Lopez 2004, pg. 9). 

(Furman and Lopez 2004, pg.13). 

• X-ray m

0006, Revision 2 (Kathren 2003). T

• For spinal x-rays, a 1.8 multiplication factor is applied to convert air kerma from a singlephase unit to a three-phase unit (Furman and Lopez 2004, pg. 10). 

To date, the records reviewed and site expert interviews only indicate standard posterio

chest x-rays. Excluding lateral chest x-rays assumed in other site profiles is reasonable for cases 

where there is no indication of such an exam in the medical records

mb r spine x-rays were included at RFP based on information indicating anterior-posterio

eral x-rays were performed at the initial em
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Th v

conside

additio ber spine x-rays upon hire from 1952–

1974 is acceptable. 

Information on x-ray machine equipment and techniques employed at RFP is unavailable until 

20 -

OTIB-0 ors at

RFP were similar to those at other DOE sites (Furman and Lopez 2004, pg. 9). In order to 

app ci

output, ess, the techniques used, and the processing equipment parameters. It is 

uncertain whether the assumption made in the RFP site profile is bounding for this particular site. 

The applic

no e

RFP, th

The IREP input criteria f

The radiation rate was assum e dose

distribu %

uncerta e or e conversion

fac s

TB . hren

2003) a site profile does not specifically give the 

-ray,

s for

ed 

sment Corporation (RAC) (Rope et al. 

e

is e aluation should consider special and/or incident related x-ray exams. This has not been

red in the RFP site profile or other site profiles reviewed to date. In the absence of 

nal information, assuming each employee had lum

01. As a result, the RFP site profile assumed the default assumptions outlined in ORAUT

006, Revision 2 (Kathren 2003). The assumption was also made that technique fact

re ate the dose received from an x-ray, it is important to know the type of equipment, the

the beam hardn

ation of the 1.8 multiplication factor between single- and three-phase units is a factor 

t us d in other site profiles reviewed. Without information on the equipment and techniques at

is factor cannot be effectively evaluated. 

or occupational medical exposure was consistent between site profiles. 

ed to be acute, the radiation type was 30-250 keV photons, th

tion was constant, and the dose was multiplied by a factor of 1.3 to account for a 30

inty. The basis for th gan dose conversion factors, the substitution dos

tor , and the analogue organs assumed in the RFP site profile were consistent with other

Ds RFP assumed the same chest wall thickness as that used in ORAUT-TIB-0006 (Kat

nd the Savannah River Site TBD. Although the

backscatter factor for skin dose, it does refer to Table B-8 in NCRP 102 (1989), Medical X

Electron Beam, and Gamma-Ray Protection for Energies up to 50 MeV (Equipment Design, 

Performance and Use). Presumably the RFP value is 1.35 as with other TBDs. 

Overall, the default values assigned for determining occupational medical exposure are relatively

consistent. Deviations from the techniques used in other TBDs are based on site specific 

information. 

Occupational Environmental Exposure 

ORAUT-TKBS-0011-4 (McDowell-Boyer and Little 2004) describes the default assumption

occupational environmental dose at RFP. The profile depends substantially on work perform

by ChemRisk (1994a, 1994b) and Radiological Asses

1999; Rood and Grogan 1999; Voillequé 1999a, 1999b, and 1999c; Weber et al. 1999). The 

purpose of these assessments was to calculate dose to members of the public, and not 

environmental dose to workers onsite. The RFP environmental site profile differs from other sit

profiles in the following ways (McDowell-Boyer and Little 2004). 

• An environmental dose is applied when “a worker was not monitored adequately to 

develop a reliable individual dose”(pg.7). 
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ient

 applied for soil resuspension calculations (pg. 11). 

• Doses are multiplied by an inhalation dose factor from ICRP 72, Age-dependent Doses to 

sed for respiratory system cancers, while soluble 

plutonium (Type M) is used for all other cases. 

orne releases from routine and “nonroutine incidents.” The site profile considered 

• The elements considered for internal environmental exposure include routine amb

airborne releases, airborne releases from the 1957 and 1969 fires, and soil resuspension 

from contamination around the 903 Pad (pg. 9). 

• A respirable fraction of 1.0 is

• The particle size is assumed to be 15 micron Activity Equivalent Diameter (AED). 

Values adopted from atmospheric modeling assumed < 30 micron AED (pg. 9-10). 

• The site profile provides a single set of dose values based on an average of onsite air 

sampling data. 

• There is no consideration of submersion dose. 

Members of the Public from Intake of Radionuclides: Part 5: Compilation of Ingestion 

and Inhalation Dose Coefficients for select radionuclides. 

• 238Pu, 241

Am, 235U, 238U, and 3H are not considered substantial contributors to the 

environmental dose, although they are known to be present at the site. There is no 

mention of whether 233/234

U and curium releases contributed to the dose. 

• Insoluble plutonium (Type S) is u

Environmental occupational dose is typically assigned to those individuals who were not 

monitored. In the case of the RFP site profile, environmental dose is applied when “a worker 

was not monitored adequately to develop a reliable individual dose.” A worker who was not 

monitored for either external or internal dose versus a worker that was not adequately monitored 

constitutes a different set of individuals. Those that were not monitored at all were likely to be 

in positions where radiation exposure was not an issue. Workers who were monitored 

inadequately can’t be compared to those who may or may not have been exposed to radiation. 

For example, if a subcontractor worked in the production area where there was a potential for 

exposure, and no dosimetry data is located for the individual, it is not appropriate to assign the 

individual an environmental dose. The terminology used for the application of occupation 

environmental dose is confusing, and should be more clearly stated to exclude inadequately

monitored workers who actually received exposure. 

Review of the site profiles to date indicates that NIOSH has not come to a consensus on what

components should be considered in the environmental dose. RFP had included dose from

ambient airb

only the 1957 and 1969 fires under “nonroutine incidents.” The RFP environmental site profile 

includes only potential exposure from soil resuspension at the 903 Pad. There are other outside

areas onsite that have soil contamination levels. In fact, the Savannah River Site (SRS) site 

profile indicates that plutonium was detected at several locations near the release points in both 

the F and H areas (Scalsky 2004, pg. 58). Soil contamination has been found at many areas on 
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ssment of internal environmental dose. This included 

itium exposure from the large tritium release in 1973. It is interesting to note that Table 4B-1 

uded data for multiple receptor points 

nsite. The RFP site profile used onsite environmental air sampling data and atmospheric 

mates based on public dose assessments to determine a single intake 

esumably, the assumption here is that the onsite air monitors will provide 

he RFP site profile has introduced a few new concepts not previously used in other site profiles. 

action of 1.0 is applied for environmental dose. This value is claimant favorable. 

e used two particles sizes: < 30 micron AED for modeled intakes and 15 micron 

E fo

size n

reconst

softwa

AMAD required under 42 CFR Part 82, unless other pa

pro

app p

The RF

Membe

Inhalat dionuclides. An inhalation dose fraction 

has not been included in other site profiles reviewed to date. It is unclear why this factor would 

apply to RFP, but not to other DOE and AWE sites. 

the Rocky Flats site, including in the East Spray Fields and the buffer zone. There is no 

explanation of why dose from radioactive material in soil excluded these other areas. SC&A was 

unable to obtain soil sample data to determine whether the 903 Pad had the highest soil 

contamination levels onsite. If this is the assumption made in the TBD, this should be clearly 

stated. 

239+240

Pu was considered the only significant source of exposure from airborne exposure in the 

environment at RFP. Other radionuclides such as 238Pu, 241Am, 235U, 238U, 3

H, 233/234U, and

curium were not considered in the asse

tr

shows no elevated tritium release during 1973. These radionuclides should be reconsidered, 

since many were purified and not merely an impurity. An evaluation of episodic releases of 

these radionuclides should be completed to verify releases from these radionuclides remain 

inconsequential. 

Site profiles such as Hanford, SRS, and INEEL have incl

o

dispersion model esti

quantity per year. Pr

cumulative release data from all operations of the plant. The site profile has not demonstrated

that this single set of intake data represents the highest potential environmental exposure onsite. 

A similar analysis at multiple receptor points is warranted to determine whether this single intake

quality is bounding for all areas onsite. 

There is no submersion dose assigned at RFP from noble gas releases. There has been no data

on large particle releases discovered to date, as with INEEL and Hanford. As a result, it is 

reasonable to exclude these components from the environmental dose. As with other site profiles

reviewed to date, there is no discussion of the liquid effluent streams. The RFP site profile also

does not consider uptake from ingestion of materials (e.g., dirt, game, vegetation, etc.). 

T

A respirable fr

The site profil

A D r intakes calculated from air sampling data. Other site profiles typically provide particle 

i terms of AMAD, as does the RFP internal site profile. There is not direction to the dose

ructor on how to convert AED to AMAD, which is the unit used within the IMBA 

re. The environmental site profile indicates that the AED values may exceed the 5 micron 

rticle size data are available. The site 

file has not presented any data to demonstrate that an alternate particle size is more 

ro riate.

P site profile introduces an inhalation dose factor from ICRP 72, Age-dependent Doses to

rs of the Public from Intake of Radionuclides: Part 5: Compilation of Ingestion and

ion Dose Coefficients (ICRP 1996) for select ra
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The

the typ per

Type S was an issue in environmental exposure scenarios. The solubility for SRS was based on 

the o

Type S lity class under unknown conditions is 

claimant favorable as was done with RFP. The RFP environmental site profile, however, did not 

con e al

for this

bounds

The assumptio th

other site profiles. There is a caveat allowing the dose reconstructor to scale numbers based on 

the work level files to 

date have cons s to ascertain external dose. The IREP input 

r radiation rate, radiation type, and dose distribution types are consistent with the radionuclides 

con e

Extern

ORAUT-TKB l

external dose at RFP. ORAUT-OTIB-0027, 

Flats Plant (Sm ons to be used in dose 

construction. etry 

pro am ose

rec st l

(NDRP

cience al 

ation. T ation has only 

recently becom een integrated into the site profile. 

The RFP external dose site profile differs from

e P

• 

the default

• 

ber of zero doses is estimated based on the 

solubility assumed for the calculation of internal dose was based on the type of cancer and

e that would result in the highest dose to the organ. There is no mention of whether Su

kn wn material in the facility. For Hanford, the solubility for plutonium was assumed to be 

. Using the most claimant-favorable solubi

sid r the presence of Super Type S material from releases to the environment. The potenti

material in the environment should be investigated to ensure current methodology 

environmental internal doses. 

ns made with respect to ventilation rate and exposure time are consistent wi

and actual length of exposure during the year. Where available, site pro

istently used environmental dosimeter

fo

sid red and the assumptions from other site profiles. 

al Exposure

S-0011-6 (Langsted 2004) describes the default assumptions for occupationa

Supplemental External Dose Information for Rocky 

ith 2005), provides additional information on the assumpti

re The assumptions were derived from historical records relating to the dosim

gr and observations made from dosimetry report review. The site profile refers the d

on ructor to the Rocky Flats work history file, the Neutron Dose Reconstruction Protoco

) file, and a job exposure matrix put together by the University of Colorado Health

S s Center and the Colorado Department of Public Health and Environment for addition

inform he job exposure matrix data is unavailable. The NDRP inform

e available to NIOSH/ORAU and has not yet b

 other site profiles in the following ways: 

B ta/ hoton 

There are two default radiation types used for plutonium exposure. ORAUT-OTIB-0027

indicates default radiation types for plutonium of 25% photons < 30 keV and 75% 

30-250 keV photons (Smith 2005, pg. 5). ORAUT-TKBS-0011-6 indicated

radiation type for plutonium of 100% 30–250 keV (Langsted 2004, pg. 25). 

Available dosimetry records do not provide individual dosimeter results for 1951–1976. 

The exchange frequency is determined by job title. If the job title is unknown, the most 

frequent exchange frequency is used. The num

dose level, the monthly, quarterly, or annual limits, and the number of possible zero 

monitoring intervals (Langsted 2004, pg. 20). 
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e default exposure geometry is based on eight major job categories and the relative 

time spent performing hands-on work for that particular job (Langsted 2004, pg. 23). 

• cords with missing entries, dose is assigned based on the preceding and 

following period (Langsted 2004, pg. 23). 

• se for plutonium workers,

mith 2005, pg. 6-7): 

he appropriate factor 

to obtain the upper 95% confidence level (Smith 2005, pg. 7). 

nt.

,

rem 

• Th

For dosimetry re

For years prior to 1960, estimation of low-energy photon do

intermediate/high-energy photon for plutonium workers, and > 15 keV electrons for 

uranium workers is calculated with the following formulas (Smith 2005, pg. 6): 

Electrons >15 keV or photon <30 keV = [skin-pen]/0.50 = Open Window (OW)

Photon 30-250 keV + photon >250 keV = pen* = Cadmium (CD) + 50%*OW

 * pen = penetrating radiation dose 

• From 1960–1970, estimation of low-energy photon dose for plutonium workers, 

intermediate/high-energy photon for plutonium workers, and > 15 keV electrons for 

uranium workers is calculated with the following formulas (Smith 2005, pg. 6): 

Electrons >15 keV or photon <30 keV = [skin-pen]/0.65 = Open Window (OW)

Photon 30-250 keV + photon >250 keV = pen = CD + Brass(BR) 

• From 1970-present, estimation of low-energy photon dose for plutonium workers, 

intermediate/high-energy photon for plutonium workers, and > 15 keV electrons for 

uranium workers is calculated with the following formulas (S

Electrons >15 keV or photon <30 keV = Skin - Pen 

Photon 30-250 keV + photon >250 keV = Pen 

• Electron and photon dosimetry uncertainty values are based on the building and 

dosimeter result range (mrem). The reported dose is multiplied by t

• For skin contamination, the location and count are obtained from personnel 

contamination reports. Prior to 1970, the exposure length is 8 hours. From 1970 

forward, the exposure length is 4 hours. The GM probe is assumed to be 33.3% efficie

Depleted uranium is aged 1 year. VARSKIN is used to calculate a dose (Langstad 2004

pg. 42). 

• The reconstructed dose for unmonitored individuals in the production area is 600 m

per year (5% of the annual Radiation Protection Guideline), or 1.2 rem per year at the 

upper 95% confidence level. 
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,

 multiplier and uncertainty factor are multiplied by the dosimeter result based 

on location and dosimeter result range (Smith 2005, pg. 7). 

rs.

to

0-250 keV. There is no explanation as to why this difference exists 

ted using Limit of Detection (LOD)/2 

dose

ttachment B (Langsted 2004, pg. 67). It is evident from

).

or

s. The MCW site profile based

to

be a clear understanding of the tasks performed. 

Neutron 

• The neutron energy distribution is based on location and includes areas in Buildings 771

776, and 707. 

• A correction

• The sensitive energy range of the NTA film is assumed to be > 800 keV. There is no 

correction factor for unmonitored energies in film after 1963. The TIB (Smith 2005,

pg. 7) recommends a 2.5 correction multiplier, while the site profile recommends a value

of 1.79. 

• The default exposure geometry is based on eight major job categories and the relative 

time spent performing hands-on work for that particular job (Langstad 2004, pg. 23). 

• There is no instruction for when to assign missed neutron dose for unmonitored worke

Based on review of several dose reconstructions, a missed neutron dose is not assigned

all claimants. The dose reconstructor is referred to the NDRP for supplemental data on 

neutron exposure. These records have recently become available, but the dose has not 

been included in many denied claims. 

The default radiation type for plutonium is stated as 25% photons < 30 keV and 75% 30-250 keV

photons in ORAUT-OTIB-0027 (Smith 2005). The default assumption in the external dosimetry

site profile is 100% 3

between the two Rocky Flats documents. This discrepancy needs to be resolved. In general, the 

radiation types are consistent with those used in other site profiles. 

Missed dose for beta, photon, and neutron dose is calcula

times the number of zero readings. A lognormal distribution with an uncertainty of 1.52 is 

applied. The difference between RFP and other sites is that the number of zeros has to be 

estimated, since only summary reports are available. This estimation is based on the annual

received, the exchange frequency for a particular job, and the annual regulatory limit. This 

approach is reasonable, assuming the job titles are categorized correctly. 

The major job titles are listed in A

worker interviews that some jobs have not been adequately characterized (see Attachment 2

For example, Clerk Packers are listed under support personnel, when their job involved handling

plutonium on a routine bases. Prior to assuming which jobs fit into which category and how 

much hands-on work was conducted by particular jobs, NIOSH/ORAU should obtain input on 

the tasks associated with these jobs. Not only may this impact the number of zeros calculated f

missed dose, but it will also affect exposure geometry selection

exposure geometry on job title as with RFP; however, the list of job titles was much more 

extensive. Hanford and SRS based their default geometry on the compensability of the claimant. 

Where possible, use of individual specific geometry factors is acceptable; however, there has
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ing 

nmonitored individual would receive a dose based on the LOD and the exchange frequency. 

le,

ium. 

anford included 20% of the open window dose in the penetrating dose. Once better designed 

acies

to

nd

explained.

Electron and photon dosimetry uncertainty values are based on location and dosimeter result 

nge (mrem). The reported dose is multiplied by the appropriate uncertainty factor to obtain the 

.

. A

r

the annual Radiation Protection Guideline), or 1.2 rem per year at the upper 95% 

confidence level. This varies from the LOD/2 times the number of zeros used in other site 

pro e

previou

favorab

The neutron energy distribution was limited to Buildings 771, 776, and 707. There was no 

spe i s were

used. T

Also, the neutron energy distribution may change when the site installed Benelex impregnated 

with boron in the glove boxes to provide additional shielding. 

In orde

into energy ranges and an ICRP 60 radiation weighting factor was applied. This is the same

RFP is the first site profile reviewed where the dose reconstructor is directed to use the preced

and following periods to assign dose for gaps in the record. In previous site profiles, an 

u

Use of the higher of the two methods is claimant favorable and should be considered. Another 

option for determining a bounding dose for a particular missing year is to use secondary 

dosimetry data. 

Two-element dosimeters did not effectively account for low-energy photons present in

plutonium facilities. This was recognized by the sites and adjustments were made. For examp

SRS used a special x-ray calibration curve for workers involved with handling pluton

H

dosimeters were introduced, this was no longer necessary. RFP also recognized the inadequ

in the two- and three-element dosimeters through 1970, when the TLD was implemented. For 

the two-element dosimeter, they added 50% of the open window dose to the penetrating dose. 

For the three-element dosimeter, they added 35% of the open window dose to the penetrating 

dose. In the RFP TIB, ORAU has backed these correction factors out of the data in order

separate the photons into energy categories. This was not done with data from Hanford a

Savannah River Site. The difference in the RFP approach has not been adequately

ra

upper 95% confidence level dose (Smith 2005, pg. 7). The consideration of dosimeter range in 

the uncertainty is important, as the closer the dosimeter is to the LOD, the higher the percentage

of error would be. 

The inclusion of a methodology for skin contamination was a positive addition to the site profile

This is especially important in facilities which had routine personnel contamination incidents

similar section should be included in other site profiles. 

The reconstructed dose for unmonitored individuals in the production area is 600 mrem per yea

(5% of

fil s. The method employed for unmonitored workers at RFP should be compared to the 

s approach of estimating missed dose based on the LOD to ensure the most claimantle methodology is used.

cif c mention of areas handling uranium, storage areas, or areas where neutron source

hese areas should also be considered when determining neutron energy distributions. 

r to calculate a neutron dose based on monitoring data, the neutron dose was separated
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me o

consist

The sensitive energy range of the NTA film is assumed to be > 800 keV. This value is consistent 

wi h

correct on film reading. The TIB (Smith 2005, pg. 7) recommends a 

2.5 correction multiplier for neutron film from 1951–1963 while the site profile recommends a 

val o in

the nex

respon

The NDRP offers a unique opportunity for NIOSH/ORAU to determine who had a missed 

neutron dose for a subset of Rocky Flats workers. The reevaluation of NTA film for plutonium

wo r

site profile. Since not all workers were included in the study, the methodology used to assign 

missed neutron dose to workers should be maintained as a part of the site profile. Clear 

instructions on when to include missed neutron dose is necessary, as there seems to be a 

discrepancy as to when it is included in a particular worker’s dose reconstruction. 

Int n

ORAU es the default assumptions for occupational internal 

dose at RFP. The assumptions we

in- ro

derived

Estima T-OTIB-0018, Internal

Do O ampling Programs (Brackett and Bihl 2005), as 

ind t RAU (see Attachment 4). Revision 0 of the internal 

dosime nmonitored workers. 

Th F es in the following

ires unless the qualifying cancer involves tissues of the extrathoracic regions 

thodology used in other site profiles. The default IREP input for neutron exposure is als

ent with other site profiles (e.g., a chronic exposure of 100 – 2,000 keV neutrons). 

th t e minimum neutron energy adopted in other site profiles. There is a discrepancy in the 

ion multiplier for the neutr

ue f 1.79 (Langstad 2004, pg. 34). NIOSH/ORAU has indicated that this will be corrected

t revision of the site profile (see Attachment 4). No correction factor for energy under 

se is applied after 1963, although TLNDs were not implemented until 1971. 

rke s serves as the basis for this. This information needs to be integrated into a revision of the

er al Exposure 

T-TKBS-0011-5 (Falk 2004) describ

re derived from historical records relating to the in-vivo and 

vit monitoring programs. The procedures used for assignment of missed internal dose are

from ORAUT-OTIB-0002, Technical Information Bulletin, Maximum Internal Dose 

tes for Certain DOE Complex Claims (Rollins 2004) and ORAU

se verestimates for Facilities with Air S

ica ed in a conference call with NIOSH/O

try site profile is incomplete with respect to internal dosimetry for u

e R P internal dose site profile (Falk 2004) differs from other site profil

ways: 

• The default solubility for plutonium, enriched uranium, and depleted uranium is type S 

for all cancers of the respiratory system and type M for all other cancers. Individuals 

involved in the October 1965 fire in Buildings 776 and 777 may exhibit super type S

characteristics (Falk 2004, pg. 8). 

• The default particle size is 5 micron AMAD. Assume 0.3 micron AMAD for all 

plutonium f

(Falk 2004, pg. 9). 

• The default isotopic composition is weapons grade plutonium. There is a separate 

isotopic composition for Zero-Power Physics Reactor (ZPPR) fuel (Falk 2004, pg. 8).
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retion period documented 

in the records (Falk 2004, pg. 11). 

DA)

rine data, instead of the Am urine data, is used to assess intakes of 

weapons grade plutonium. The 241Am dose is calculated based on a measured or assumed 

0.0036

• An index of 1.35 should be assumed to determine the MDA for lung counting, if the 

for all

r type is

r S should not be limited to the 1965 fires, but the dose reconstructor 

f the

ific 

o a number of site profiles. RFP is no exception. There have been 

article size studies completed at Rocky Flats that indicate particle sizes on the order of 1–

• Assume dpm/24-hours through 1989 and dpm/sample after 1989, regardless of the 

volume. Excretion period can be modified based on actual exc

• Uncertainty is estimated by dividing the median Minimum Detectable Activity (M

value by 3.3 (pp. 13–16). 

• The plutonium u

241

241Am concentration for the plutonium mixture (pg. 13). The initial 241

Pu mass fraction 

for the 1950s through 1976 was 0.005. July 1976–1989, the 241Pu mass fraction is

(Falk 2004, pg. 20). 

• For the purposes of lung counts, assume an initial concentration of 100 ppm 241Am (Falk 

2004, pg. 20). 

worker height and weight are unavailable (Falk 2004, pg. 19). 

• If there is residual plutonium at a wound site, an acute injection plus a possible long-term

chronic injection should be considered. For uranium contaminated wounds, consider an 

acute injection (Falk 2004, pg. 22). 

• A hypothetical intake can be calculated using ORAUT-OTIB-0002 or ORAUT-OTIB0018 where there was an established air monitoring program. When using the ORAUTOTIB-0002 methodology, the 28-radionuclide scenario is used. 

• No missed tritium dose is assigned. 

• There is no consideration for ingestion dose in monitored workers. 

The default solubility for plutonium, enriched uranium, and depleted uranium is type S

cancers of the respiratory system and type M for all other cancers. Individuals involved in the

October 1965 fire in Buildings 776 and 777 may exhibit super type S characteristics (pg. 8). 

Assigning the most claimant-favorable solubility for a particular radionuclide and cance

appropriate. Supe Type

should consider operations and incidental fires that may have heated plutonium to temperatures 

in excess of 600 degree Celsius. High-fired uranium oxide and associated particle sizes should 

also be addressed. 

The default particle size is 5 micron AMAD. The site profile indicates that a 0.3 micron AMAD

for all plutonium fires may be considered unless the qualifying cancer involves tissues o

extrathoracic regions (pg. 9). NIOSH/ORAU has been hesitant to integrate facility-spec

particle size studies int

p
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n

ing

ction

n, the approach in the Hanford site profile is more 

laimant favorable. 

iation

es 

ulated 

ne data.

rded separately, multiply the Pu result by 1.0264 (pg.11). This is similar 

methodologies used in other site profiles to ascertain dose contribution for those radionuclides 

he

nd site.

initial concentration of 100 ppm Am (pg. 20). 

This is a reasonable assumption, as long as records don’t indicate material had a lower 

erson. As a best estimate 

pproach, using 1.35 is reasonable; however, if this data is used to calculate a maximizing dose, 

the index giving the highest MDA for a detection system should be used to be claimant 

vorable. The application of the index in conjunction with the MDA provides a more 

dividualized MDA.

2 micron AMAD. NIOSH/ORAU is obligated to consider these studies in their analyses whe

the values are more claimant favorable than the default assumptions. 

The default isotopic composition is weapons grade plutonium. There is a separate isotopic 

composition for ZPPR plutonium (pg. 8). It is appropriate to consider alternate isotopic 

compositions that workers may have encountered at the facility. The default isotopic 

composition for enriched uranium and depleted uranium is consistent with other sites handl

uranium of this type. 

In the RFP site profile, it was assumed that the sample volume represented a 24-hour colle

and no corrections were made to the volume, whether they represented an actual 24-hour 

collection or not. In the Hanford site profile, it is recommended that a urine volume of less than

reference man or woman be normalized to these values. A consistent method of addressing 

sample volume is needed. Ultimately, determining individual 24-hour sample volumes is ideal.

In the absence of true 24-hour sample collectio

c

Per the Internal Dose Reconstruction Implementation Guide (OCAS 2002b), a standard dev

of 0.3 times the MDA or reporting level, is adequate except for chest counts for which 0.5 tim

the MDA should be used. Dividing the MDA by 3.3 essentially gives the same value. The 

actual standard deviation or error is preferable where available, if it exceeds the value calc

above. 

The RFP assumes 239+240

Pu as the radionuclide of concern for nonspecific plutonium uri

If 238Pu is not reco 239+240

to

not routinely evaluated in the bioassay program. 

Although internal uptakes by wounds have been an issue at other sites, the RFP site profile is t

only site profile reviewed to date that provides direction on assessing uptakes via a wou

Consideration is given to acute and long-term chronic injection as a mode of intake. 

For the purposes of lung counts, assume an 241

concentration of 241Am. 

An index of 1.35 is assumed to determine the MDA for lung counting, if the worker’s height and

weight are unavailable (pg. 19). Where height and weight are available, the index can be 

calculated and the most appropriate MDA applied to the particular p

a

fa

in
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In a con g

approach for mis ORAUTOTIB-0002 (Rollins 2 gh RFP is 

not a reactor facility, the 28-radionuclide intake scenario for facilities with reactors was used. 

NIOSH/ORAU has provided no rationale for using the hypothetical intake for a reactor versus 

non-reactor facility. It is important to note that the internal dosimetry site profile does not 

mention the use of the above procedures for determining a maximum dose to unmonitored 

workers. There is also no explanation for when to use which TIB method. 

There has been no consideration of ingestion dose from workers at the RFP facility. The 

Bethlehem Steel and Mallinckrodt Chemical Works site profiles have included dose from

ingestion of radioactive material. Although there may have been more engineering controls at 

RFP, failure of these engineering controls, internal contamination of respirators, incidents, and 

allowing food and beverages in specific areas of the plant (e.g., uranium processing area) may 

have led to ingestion of radioactive material. 

The inputs to IREP include radiation rate, radiation type, and dose distribution type. The RFP 

inputs into IREP are consistent with the other DOE sites reviewed so far. The maximizing 

approach assumes an uptake of 28 radionuclides as defined in ORAUT-OTIB-0002 (Rollins 

2004). The exposure rate is inputted as a chronic intake. The radiation types include photons 

(>250 keV), electrons (> 15 keV), and alpha. These values are entered as a constant and 

encompass the uncertainty related to the dose estimation. The input for the best estimate internal 

dose is dependent on the individual’s internal exposure history, but will default to a chronic 

intake. It is noted in compensable cases that dose reconstructors are using a triangular 

distribution. 

Although radionuclides such as 237

Np, 238Pu, 244Cm, 233U, and thorium are mentioned as either 

trace radionuclides or purified radionuclides, there has been no consideration of potential intakes 

from these radionuclides. There is also no mention of whether missed tritium dose should be 

assigned under any conditions. The site profile indicates that dosimetry data for these 

radionuclides is rare or not available. There is no clear conclusion on whether the gross alpha 

analysis of urine included these radionuclides. The relative contribution of the radionuclides to 

internal dose is not adequately discussed. 

The Rocky Flats site profile has included example dosimetry records and detailed explanation of 

these records in both the internal and external site profiles. This is extremely helpful to the dose 

reconstructor in the interpretation of records and thus in the dose reconstruction effort. To date, 

other site profiles have not included this type of discussion; however, SC&A would strongly 

recommend that this information be included in other site profiles. 

ference call with TBD authors (Attachment 4), it was indicated that the maximizin

sed internal dose to unmonitored workers is determined by using

004) or ORAUT-OTIB-0018 (Brackett and Bihl 2005). Althou
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ATTACHMENT 6: PRELIMINARY TALKING POINTS FOR 

CONFERENCE CALL – SEPTEMBER 6–8, 2005 

9/4/2005 1

Review of the NIOSH Site Profile

for the 

Rocky Flats Site

Preliminary Talking Points for

Conference Call – September 6-8, 2005

SC&A Review Team

9/4/2005 2

Site Description: Issues and findings

Overall, the Site Description provides a reasonable characterization 

of the history of production. (1951-89). However there remain 

important gaps:

• (1) Cleanup mission, now in its 15th year at Rocky Flats, is not

included in the TBD.

• (2) Excess PU and HEU Storage vulnerabilities are not addressed.

• (3) Missing and incomplete information relative to extensive

processing of recycled uranium.

• (4) Inadequate description of U-233 processing activities
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9/4/2005 3

External Dose TBD: Findings and Issues

Records/Methodology

• Dose reconstruction from badge readings versus historical documents:

3⁄4 The TBD does not clearly relate the use of OW, CD, and BR to the

historical records’ definitions of skin, penetrating, and hand dose.

3⁄4 Dosimetry method at time of records are not always identifiable.

• Incomplete or inconsistent information:

3⁄4 1952-1975 -> the deep gamma dose and the neutron dose were 

recorded as the total penetrating dose and not recorded as separate

doses in the worker’s file.

3⁄4 Potential external doses from recycled uranium not considered.

3⁄4 1976, a large number of neutron and gamma doses were erroneously

recorded.

3⁄4 July 1984 to Oct 1984, some neutron dose was recorded, but the 

gamma and total dose was zero. 

3⁄4 1984-1986 Possible manual corrections were needed, but the 

information was not provided to make these corrections.

9/4/2005 4

Dose Calculations

External Dose TBD

• Job exposure matrix study should be available and used.

• Doses <10 mrem counted as zero dose is inconsistent.

• Use of “activity date” is uncertain.

• Worker’s badge calibrated for one work location. However,

3⁄4 Workers worked at temporary sites.

3⁄4 Workers worked at other sites on overtime assignments.

This could lead to under-recorded doses.

• Exposure Geometry – The assignment of ISO or ROT instead of AP

geometry may not always be claimant favorable.

• NIOSH needs a consistent approach for angular dependence of

beta/gamma film, NTA Film, and TLDs.
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9/4/2005 5

External Dose TBD

External Photon and Beta Dose

• Photons with E > 250 keV:

3⁄4 Penetrating power of photons w/ E>250 KeV may off set the 

REF of lower energy photons.

• Use of VARSKIN Software:

3⁄4 It is not apparent that this program includes one of the 

predominate beta emitter, Pa-234m, that was present at RFP.

9/4/2005 6

External Dose TBD

Extremity Dose

• Considerable hands-on work at RFP

3⁄4 No apparent extremity neutron monitoring.

3⁄4 Extremity dose should not be assumed to be equal to whole 

body dose:

9Extremity dose may be 6x to 12x WB dose.

3⁄4 No valid hand-to-wrist ratios provided:

9Surface beta dose rates on the order of 1 to 20 rad per hours

in some locations.
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9/4/2005 7

External Dose TBD

Neutron Exposures and Dosimetry

• Use of neutron track plates and uncertainties in neutron dosimetry

1951-1957 leads to areas of concern in correct DR:

3⁄4 Relationship of “plates” to NTA film not known

3⁄4 Characteristics/calibration of plates unknown.

3⁄4 Details of 1956-57 out-source to HPS are not known.

• NTA film neutron energy threshold may have lead to missed dose in

RFP records:

3⁄4 Early threshold of 0.25 -0.4 MeV is too low.

3⁄4 Later threshold of 0.48 – 0.75 MeV is also too low.

9/4/2005 8

External Dose TBD

Neutron Exposures and Dosimetry

• The NDRP report is limited:

3⁄4 NDRP does not cover non-Pu workers.

3⁄4 Is not directly applicable to non-Pu workers.

• Because the NDRP report does not include non-Pu workers:

3⁄4 More work is needed on the use of neutron/photon ratios and

film/TLD comparisons to correctly determine past neutron doses.

3⁄4 More robust neutron dose multiplication factors are needed.

3⁄4 Shift in neutron/photon ratios after production period not 

addressed
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9/4/2005 9

External Dose TBD

Unmonitored Individuals

• Early radiation hazards may not have been recognized before 

exposures occurred:

3⁄4 Policy of badging only those expected to receive >10% of RPG 

may have left some radiation workers unbadged.

3⁄4 Lack of information on criticality parameter controls which 

evolved over time.

3⁄4 The NDRP (Falk 2005) reports that:

9There were significant neutron fields from the alpha-neutron 

reaction with light elements, especially from the plutonium

tetrafluoride compounds.

9Pu workers were not routinely monitored for neutron

exposures until the 1960s.

9/4/2005 10

External Dose TBD

Unmonitored Individuals

3⁄4 1959-1969 -> RFP 1984 annual report and NDRP 2003 shows

that:

9Average of only 25% of workers badged received NTA film

during this 11-year period.

9Range of 9% - 50% over the 11-year period.
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9/4/2005 11

External Dose TBD

Industrial X-ray Units and Neutron Generators

• Radiographic X-ray units and neutron generator units were

apparently used for NDT at the RFP

3⁄4 Units were not addressed in TBDs.

3⁄4 Energy/current/type, periods of operation, and SOPs, of these

units need to be determined. 

3⁄4 Dosimetry systems’ response to these different radiation sources

needs addressed.

3⁄4 Possible accidents/incidents need to be investigated.

9/4/2005 12

Internal Dose TBD: Findings and Issues

• The internal dosimetry TBD is incomplete:

3⁄4 There is limited direction to the dose reconstructor on the

process and assumptions that should be used to calculate

internal dose.

3⁄4 TBD does not provide guidance for assessment of missed dose

and unmonitored workers.

3⁄4 The approaches regarding to solubility and particle size need to

be reviewed.

3⁄4 TBD does not provide guidance for assessment of dose in 

accident case.
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9/4/2005 13

Source term:

3⁄4 The TBD does not present a guidance for dose reconstruction to

associated with tritium, neptunium, thorium, curium and fission

products.

3⁄4 The TBD does not address exposures from the processing of

recycled uranium.

Internal Dose TBD

9/4/2005 14

Internal Dose TBD

MDA of bioassay data:

3⁄4 The MDA values for urinalysis are high, especially in the early years. 

3⁄4 The TBD does not present a guidance to assess the missed dose. The

graphic below illustrates an example of missed dose for one year inhalation

Pu-239, Type S, based on the different values of median MDA for plutonium

in urine.

One Year Committed Equivalent Dose for the Median MDA

Values for Pu-239

1.00E-07

1.00E-06

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

Committed equivalent dose Sv/Bq in

urine

0.57

0.54

0.51

0.44

0.24

0.02
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]

9/4/2005 15

• Particle size:

3⁄4 The TBD does not present a complete analysis of the particle

sizes. It adopts the default parameter of 5μm AMAD. 

3⁄4 In the fire accidents the AMAD values were evaluated as 0.12 -

0.30 μm.

Internal Dose TBD

9/4/2005 16

Internal Dose TBD

• Solubility:

3⁄4 The TBD assumption for the solubility of the compounds is not claimant

favorable, especially for the cancers in the gastrointestinal organs. The

solubility should be assigned based on the type of cancer as

recommended in the 42CFR Part 82.

One Year Committed Equivalent Dose for Pu-239 (Sv per Bq

excreted in the 24hr working day urine samples after 1 year of

work)

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

Adrenals

Bladder

BoneBrain

Breasts

Esophag

St Wall

SI Wall

ULI Wall

LLI Wall

Colon

Kidneys

Liver

Muscle

Ovaries

Pancrea

Red

Skin

SpleenTestes

Thymus

Thyroid

Uterus

Systemic organs

Committed Equivalent Dose

(Sv/Bq in urine)

Inhalation Type M

Inhalation Type S
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Internal Dose TBD

• Fire accidents:

3⁄4 The Table shows the similarity of the values of the IRF for urine and feces for Type S and high fired Pu compounds. It can be considered as Type S compound.

3⁄4 The urinalysis were probably inappropriate to detect the intakes of either high-fired Pu or Type S Pu, unless there was a very high intake.

3⁄4 Feces analysis seems to be the more appropriate to evaluate accidental intakes.

Dose coefficients for Pu-239 inhalation

90 6.74E-07 2.20E-07 3.83E-07 1.24E-07 4.93E-04 4.94E-04 2.77E-04 2.78E-04

30 6.61E-07 3.02E-07 3.76E-07 1.70E-07 1.52E-03 1.52E-03 8.47E-04 8.46E-04

10 7.07E-07 7.39E-07 4.05E-07 4.16E-07 2.37E-03 2.36E-03 1.33E-03 1.32E-03

9 7.26E-07 9.12E-07 4.17E-07 5.13E-07 2.46E-03 2.45E-03 1.39E-03 1.39E-03

8 7.53E-07 1.16E-06 4.34E-07 6.54E-07 2.61E-03 2.61E-03 1.52E-03 1.51E-03

7 7.92E-07 1.53E-06 4.59E-07 8.60E-07 2.94E-03 2.94E-03 1.80E-03 1.80E-03

6 8.49E-07 2.07E-06 4.95E-07 1.17E-06 3.72E-03 3.71E-03 2.52E-03 2.51E-03

5 9.35E-07 2.91E-06 5.50E-07 1.64E-06 5.60E-03 5.59E-03 4.31E-03 4.31E-03

4 1.07E-06 4.18E-06 6.37E-07 2.35E-06 9.91E-03 9.91E-03 8.63E-03 8.63E-03

3 1.33E-06 5.99E-06 7.99E-07 3.37E-06 1.82E-02 1.82E-02 1.77E-02 1.77E-02

2 1.90E-06 7.77E-06 1.15E-06 4.37E-06 2.66E-02 2.66E-02 3.01E-02 3.01E-02

1 2.96E-06 5.05E-06 1.83E-06 2.84E-06 1.33E-02 1.33E-02 1.84E-02 1.84E-02

Type S High fired Type S High fired Type S High fired Type S High fired

0.12 um 0.12 um 0.3 um 0.3 um 0.12 um 0.12 um 0.3 um 0.3 um

24 hr Urine (Bq/Bq intake) 24 hr feces (Bq/Bq intake)

days after 

intake
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Internal Dose TBD

• Fire accidents:

3⁄4 The Table shows the

similarity of the dose

coefficients for Type S

and high fired Pu

compounds. 

Uterus 4.65E-09 2.22E-09 2.65E-09 1.25E-09

Thyroid 4.65E-09 2.22E-09 2.65E-09 1.25E-09

Thymus 4.65E-09 2.22E-09 2.65E-09 1.25E-09

Testes 1.99E-08 9.90E-09 1.14E-08 5.58E-09

Spleen 4.65E-09 2.22E-09 2.65E-09 1.25E-09

Skin 4.65E-09 2.22E-09 2.65E-09 1.25E-09

Lungs 9.93E-05 9.95E-05 5.42E-05 5.44E-05

ET Airways 2.20E-06 2.22E-06 3.81E-06 3.86E-06

Red Marrow 1.99E-07 9.86E-08 1.14E-07 5.55E-08

Pancreas 4.65E-09 2.22E-09 2.65E-09 1.25E-09

Ovaries 1.96E-08 9.73E-09 1.12E-08 5.48E-09

Muscle 4.65E-09 2.22E-09 2.65E-09 1.25E-09

Liver 3.26E-07 1.62E-07 1.86E-07 9.16E-08

Kidneys 4.58E-08 2.00E-08 2.61E-08 1.13E-08

Colon 1.22E-08 9.81E-09 8.18E-09 6.79E-09

SI Wall 5.33E-09 2.91E-09 3.15E-09 1.75E-09

St Wall 4.92E-09 2.50E-09 2.85E-09 1.45E-09

Esophagus 4.65E-09 2.22E-09 2.65E-09 1.25E-09

Breasts 4.65E-09 2.22E-09 2.65E-09 1.25E-09

Brain 4.65E-09 2.22E-09 2.65E-09 1.25E-09

Bone Surface 1.96E-06 9.70E-07 1.12E-06 5.47E-07

Bladder Wall 4.65E-09 2.22E-09 2.65E-09 1.25E-09

Adrenals 4.65E-09 2.22E-09 2.65E-09 1.25E-09

Type S High fired Type S High fired

0.12 μm 0.3 μm

Organs

Dose coefficients for Pu-239 inhalation

9/4/2005 19

Internal Dose TBD

Internal Dosimetry Records/Methodology

Urinalysis

• Gross alpha urinalysis was used for workers who were potentially

exposed to uranium, plutonium and other alpha emitters in the same

monitoring period. Until 1963, the results were assigned to EU. After

1963 (and EU operations were phased out), the results were

assigned to plutonium. It is not a claimant favorable approach since

the dose for plutonium is higher than the ones for uranium.

• High uncertainties on estimates of the: high blank, recovery, 24

hours urine volume, and consequently on the estimate of the median

MDA values. The TBD presents the MDA for extreme condition, but

it does not provide information on how to apply them. 
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Internal Dose TBD

Internal Dosimetry Records/Methodology

Urinalysis

• Urinalysis results less than 10% of the tolerance level were recorded

and reported as background (BK on the Urinalysis Record Card) or

zero – it is inconsistent and critical, especially for alpha emitters.

• There is no clear guidance on how to normalize the urine sample

volume for 24 hours urine volume.

9/4/2005 21

Internal Dose TBD

Internal Dosimetry Records/Methodology

In Vivo Lung Count

• Assignment of the date of intake – source of uncertainty.

• It is not clear if in vivo lung count was used as a routine monitoring

program or just in special monitoring.

• No information about measurement of fission products.

• A positive detection of Am-241 did not necessarily indicate an intake

of the plutonium/americium mixture, especially for a lung count in

response to an incident – it is not a claimant favorable approach.

• Consider super-equilibrium of the Th-234 with the U-238 is not

necessarily claimant favorable approach, especially for the workers 

exposed to depleted uranium metal with a deficiency of Th-234.
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Internal Dose TBD

Internal Dosimetry Records/Methodology

In Vivo Lung Count

• The assumption of the similar behavior of americium associated with 

the plutonium particles in the lungs until the particles are dissolved

or removed from the lungs – it need to be evaluated in order to avoid

the underestimation of plutonium activity. 

• When there is no measurement in one of the chest side, it is stated

that the dose reconstructor should estimate the contribution for the

other chest side before using data from the count – no guidance for

this calculation.

9/4/2005 23

Internal Dose TBD

Other Bioassay Data

• Wound count information is largely irrelevant to dose reconstruction:

3⁄4 Although no appropriate wound model is available currently, the

approach is claimant favorable for most types of cancer, unless for 

lymph node cancers – special approach need to be presented, since

there are available information about a signiticant fraction of intake

retained in the lymph nodes.

• Fecal analysis: 

3⁄4 The available data should be used as much as possible to evaluate the

intakes of radionuclides Type S compounds.
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• Potential for Ingestion Exposure Pathways:

3⁄4 Ingestion pathway of exposure was not considered on the dose

assessment. The ingestion of insoluble compounds should be

included as a part of the internal dose assessment, especially in

cases of cancer in organs of the gastrointestinal tract.

Internal Dose TBD: Dose Assessment

One Year Committed Equivalent Dose for Pu-239 (Sv per

Bq excreted in the 24hr working day urine samples after 1 

year of work)

1.00E-03

1.00E-02

1.00E-01

1.00E+00

1.00E+01

1.00E+02

1.00E+03

Adrenals

Bladder

BoneBrain

Breasts

Esophag

St Wall

SI Wall

ULI Wall

LLI Wall

Colon

Kidneys

Liver

Muscle

Ovaries

Pancrea

Red

ET

Lungs

Skin

SpleenTestes

Thymus

Thyroid

Uterus

Organs

Committed Equivalent

Dose Sv/Bq in Urine

Inhalation Type M

Inhalation Type S

Ingestion f1=0.0005

Ingestion f1=0.0001

Ingestion f1=0.00001

9/4/2005 25

Internal Dose TBD: Dose Assessment

• Coworker Dose Assignment:

3⁄4 The TBD does not present any guidance to calculate the dose

for unmonitored workers.
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Occupational Medical Exposure TBD: Findings and

Issues

• NIOSH needs to better define its interpretation of what 

constitutes medical exposure.

3⁄4 NIOSH does not provide clarity or interpretation of included 

exposures in OCAS (2002) guidelines.

3⁄4 ORAUT-TKBS-0011-3 interprets NIOSH guidelines as including 

pre-employment and annual routine exams.

3⁄4 Special x-rays (e.g., respiratory certification) are presumed to

have occurred at time of annual medical exam, which is not 

demonstrated by records.

3⁄4 ORAUT-OTIB-0006 recently revised (Revision 3) suggests that

all exposure incidental to employment should be included. TIB 

also suggests inclusion of termination x-rays be included.

3⁄4 TBD does not discuss impacts of special exams and x-rays

resulting from work related injury or illness, and routine-to-work

exams.

9/4/2005 27

Occupational Medical Exposure TBD

• The TBD does not address the potential for

sealed sources being used in the medical clinic. The TBD does not catalog the number and types 

of x-ray equipment available for use and the

conditions of use.

3⁄4The TBD provides little or no information on types of

procedures used on workers or protocols for use 

beyond chest x-rays.

3⁄4The TBD does not detail the protocols used to assure

workers did not receive x-ray exams without 

approvals or received unwarranted screening exams.

3⁄4The TBD does not document that x-ray equipment did

not exceed the diagnostic range or greater than 80

kVp.
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Occupational Medical Exposure TBD

• The TBD uses assumptions drawn from other site

reviews to address exam frequencies. The TBD

assumes only 1-2 exams per year. This is a very

speculative assumption for many workers without proper

documentation.

3⁄4 There is no documentation to show that most workers did not 

receive voluntary annual chest x-rays.

3⁄4 There is no documentation to show that special worker

categories (e.g., asbestos and beryllium workers) did not receive 

multiple chest x-rays.

3⁄4 The RFP utilized the PFG unit for nearly 20 years (~1950-1970)

which have the highest exposure rates.

3⁄4 The TIB suggests that a retake rate of 3% was used. The use of

Type II machines in industrial environments may easily run an

order of magnitude higher (up to 30%).

9/4/2005 29

Occupational Medical Exposure TBD

• There is little documentation to support the assumption 

on equipment and x-ray exam techniques which appear

to be derived mainly from TIB-0006.

3⁄4 There is no documentation on x-ray protocols or machine 

calibration requirements prior to 2001.

3⁄4 Their assumptions rely heavily on ORAU 2003a and b from the 

SRS.

3⁄4 Their dose to abdominal and spinal areas are derived from

Lincoln and Gupton (~1958).

3⁄4 The RFP relied heavily on the PFG unit for routine chest 

radiography. Doses to workers could have been easily 5-6 times

as great as other sites from chest x-rays alone.

3⁄4 The TBD does not address the importance of processors which 

impact dose and retakes.
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Occupational Medic

t evaluate the dose impact due

to proper use of screens or grids, as it assumes 

r impact 

off-site exposures.

e is no documentation to show that the RFP 

ve been prescribed by

cians at RFP.

al Exposure TBD

• The TBD does no

that equipment output would not vary o

3⁄4Ther

assessed its techniques and protocols to minimize

dose.

3⁄4The RFP TBD does not address off-site medical

exposures that may ha

physi

3⁄4There are no records to show that physicians 

reviewed the frequency, type or need for x-ray exams 

on a routine basis.

Occupational Environmental TBD: Findings and Issues

3⁄4 Values applied in 1965–2002 dose reconstruction.

3⁄4 Identify if th

9/4/2005 31

e 50th or 95th percentile values were used to determine dose

3⁄4 Why ingestion of Pu and Am was not included in the dose calculations

uacy of RATCHET to simulate the atmospheric dispersion factors on

relevance

3⁄4 The relationship and relevance of particle sizes need to be explained

3⁄4 Many particle sizes were used in the text and relevance to dose was not clear

3⁄4 Inadvertent ingestion of radioactive materials

3⁄4 Continuous resuspension of plutonium and americium 

3⁄4 Resuspension seemed ignored in calculating the dose

3⁄4 Adeq

site

3⁄4 Code was originally designed for off-site calculations

3⁄4 Delineate phases of operations, data availability, and types of data. 

3⁄4 Several Phases were being identified and time-line is needed to emphasize 
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A

DOCUMENTS 

The those that need to be explained or addressed further in order to provide 

clar

Observation 1: Errors or Co

issing Information

TTACHMENT 7: EDITORIAL ERRORS IN THE TECHNICAL BASIS

following items are

ity and/or completeness in the TBD. 

nfusing Statements in TBD-0011-6 

M

RAU-TKBS-0011-6 (Langsted 2004, pg. 18), 3rd line, lists, “1977-present, dosimeter 

exchange history.” Shouldn’t there be more information provided besides the exchange 

history (i.e., quarterly dose history, etc.)? 

Energy Range Error

O

ORAU-TKBS-0011-6 (Langsted 2004, pg. 24), 6th line up from the bottom, should read ...the 

30-250 keV photon..., not ...the 30-50 keV photon.... 

Table 6-18 Clarification

ORAU-TKBS-0011-6 (Langsted 2004, pg. 31) Table 6-18, Column 2 heading, should read 

“Fraction of dose using NCRP38 flux-to-dose conversion factors.” This would clarify how 

the information in the column was derived. Also, potential missed neutron dose based on 

accompanying text should be 60% with a multiplying factor of 2.5 rather than 56% and 1.79 

from the standpoint of claimant favorability. 

Table 6-24 Clarification

ORAU-TKBS-0011-6 (Langsted 2004, pg. 41) Table 6-24 footnotes state to multiply any 

dose greater than 2 mrem by 1.12. However, the fourth column contains numbers that do not 

use this multiplier and do not match Columns 2 and 3. 

Observation 2: Errors or Confusing Statements in OTIB-0027 

Reference Error

ORAU-OTIB-0027 (Smith 2005), Page 5, 1st paragraph, 3rd hird line, should read, Rocky 

Flats Plant – Occupational Environmental Dosimetry (Rev 01).

2 [Emphasis added.] 

Error in Calculating Skin and Penetrating Doses

ORAU-OTIB-0027 (Smith 2005), page 5, in the Pre-1960 section list that skin dose was 

determined by (OW + CD + BR) and also by (OW + CD). The BR term is apparently in 

error in this paragraph and should be deleted. 

Beta/gamma only in OTIB-0027

ORAU-OTIB-0027 (Smith 2005) provides supplemental DR information for beta and gamma 

doses only. It would be helpful to provide in the document title the fact that it is only for 

beta/gamma (i.e., it does not apply to neutron dose reconstruction). This would clarify its 
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Table 4-2 on pages 7 and 8. 

purpose, because it does mention neutrons in the uncertainty factors listed in Table 4-1 and

Error in Table 4-2

ORAU-OTIB-0027 (Smith 2005), page 8; in Table 4-2 it would appear that the third entry in 

Column 2 should read 1983-2003 instead of 1983–1998. 

 
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ATTACHMENT 8: RECORDS REQUESTED FROM ROCKY FLATS 

Rocky Flats records personnel performed a keyword search of their site records database for the 

SC&A team. The list of records under each keyword was reviewed for technical reports and 

other records pertinent to the SC&A review. The following records were requested from RFP 

via email on September 9, 2005. Records have not been provided to date. 

(1) Investigation Team Report: Investigating the Source of Potential Internal Radiological 

Exposures Involving Eleven Personnel in Building 771, Rocky Flats Environmental 

Technology Site, March 15, 2001. 

(2) Sullivan, M.T., Lifetime Dose Limitation and Neutron Dose Reconstruction, Letter, July 

26, 1994. 

(3) Tontodonato, R.E., Trip to Review Feed Characterization for RFP Building 707 Thermal 

Stabilization Process, January 20, 1994, Letter, February 8, 1994. 

(4) A Report on Radiation Problems Related to Plutonium Fabrication Operations at the 

Rocky Flats Plant, Dow Chemical, Rocky Flats Division, January 1968. 

(5) Safety Considerations in the Operations of the Rocky Flats Plutonium Processing Plant,

Atomic Energy Commission, March 1970. 

(6) Putzier, Edward, 1970, A Summary of On-site Radioactive Waste Disposal, April, 1970. 

(7) Putzier, Edward, Past Thirty Years, 

(8) Barrick, C.W., 1981, Past Accidental Releases of Radioactivity from the Rocky Flats

Plant, January 14, 1981. 

(9) Putzier, Edward, Internal Letter to Medical File, March 8, 1976. 

(10) An Aerial Radiological Survey of the United States Department of Energy’s Rocky Flats 

Plant, DOE, August 1981. 

(11) An Aerial Radiological Survey of the United States Department of Energy’s Rocky Flats 

Plant, DOE, July 1979. 

(12) Environmental Safety and Health Progress Assessment of the Rocky Flats Plant, DOE, 

May 1993. 

(13) Measurements of Plutonium-238, 1967. 

(14) In-vivo Measurements of Plutonium-238, 1967 (Bistline, pg. 6). 
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(1

Control of Sealed Radioac tment of Energy Sites." 

5) 2002-03-12 – Report DOE/IG-0544: Report on "Inspection of the Accountability and 

tive Sources at Selected Depar

(16) 1996-09-19 – Report WR-L-96-03: Audit of the Environmental Restoration of the 903 

Pad, Mound, and East Trenches at the Rocky Flats Environmental Technology Site. 

(17) 1995-03-15 – Report WR-L-95-22: Assessment of EG&G Rocky Flats, Inc., Internal 

Audit Function. 

(18) List of Site Closure Reports in the RFP Records System. 
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AT S.

The absorp the in hysic l form. 

ICRP recommends that material-specific rates of ab should be use

compounds for which reliable experimental data ex ther compounds, default values of 

param recommended. In Rocky Flats, high-fired oxides were generated during the two 

big fire accidents. It is clear r potential so uch oxides are numerous 

smalle ave occurred, and multiple high temperature processes in furnaces, 

incine roduction reas used in m lutonium build

In order to evaluate the effec f bioassay m hich were applied during the early 

years y Flats, to dete ctivity in urinary samples from accidental intakes, some

simulations of different scenarios for 239Pu inhalation were done. As ICRP

recom specific value retention pa or high-fired p , the Type S 

param ere used for the tions. The Type S plutonium compounds are less 

solu n high-fired plu not possible to detect measurable activity for 239Pu 

Scenario 1 - Chronic inhalation different compounds: For the 239

Pu activity of 1 dpm/24-hr 

in urine samples taken at the end of the 20 years of exposure, assuming that the total activity is 

due to the inhalation of Type S compound (AMAD = 1μm), the total intake was calculated to be 

6.42E+04 Bq. Assuming that the total activity was due to the inhalation of Type M compound 

(AMAD = 1μm), the total intake was calculated to be 6.68E+03. The equivalent doses are 

presented in Table 1, which shows that the doses are higher for Type M compound compared to 

the insoluble ones, except where the critical organs are the large intestine and respiratory tract, 

where the high-fired compound delivers the highest doses. 

Table 1. Equivalent Dose due to a Chronic Inhalation of Plutonium-239 for 

Type S and Type M Compounds, Estimated by Urinalysis 

Equivalent dose for 20 y chronic exposure (Sv) 

TACHMENT 9: EVALUATION OF INHALATION SUPER S V

TYPE S AND TYPE M 

tion of haled material to the blood depends on its p al and chemica

sorption d in the model for 

ist. For o

eters are

that othe urces of s the more

r fires that h

rators, and p process a ultiple p ings.

tiveness o ethods, w

in Rock ct 239

Pu a

has not

mended s for lung rameters f lutonium

eters w

ble tha

se simula

in tonium, so if it is 

Type S in urinary samples due to the high MDA values it will be unlikely to measure any activity

in urinary samples due to acute intake of high-fired plutonium. 

Organs Type S Type M TypeM/TypeS

Adrenals 3.28E-03 4.15E-03 1.27E+00

Bladde

one Surface 1.95E+00 2.52E+00 1.29E+00

Brai 3.28E-03 1.27E+00

Brea 4.15E-03

agus 4

ll 4.17E-03 

all 4

all 4.19E-03 

all 4

4.22E-03 

ys 1

r Wall 3.28E-03 4.15E-03 1.27E+00

B

n 

sts

4.15E-03

3.28E-03 1.27E+00

Esoph 3.28E-03 .15E-03 1.27E+00

St Wa 3.30E-03 1.26E+00

SI W 3.34E-03 .17E-03 1.25E+00

ULI W 3.66E-03 1.14E+00

LLI W 4.43E-03 .26E-03 9.62E-01

Colon 3.99E-03 1.06E+00

Kidne 1.50E-02 .82E-02 1.21E+00
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Equivalent dose for

Table 1. Equivalent Dose due to a Chronic Inhalation of Plutonium-239 for 

Type S and Type M Compounds, Estimated by Urinalysis 

20 y chronic exposure (Sv) Organs Type M 

5

Type S TypeM/TypeS

Liver 4.04E-01 .28E-01 1.31E+00

Muscle 4.15E-03 

s 3

as 4.15E-03 

arrow 1

ways 13E-02 

1

4.15E-03 

4

3.23E-02 

4

d 4.15E-03 

4

3.28E-03 1.27E+00

Ovarie 2.42E-02 .16E-02 1.31E+00

Pancre 3.28E-03 1.27E+00

Red M 1.50E-01 .88E-01 1.25E+00

ET Air 2.44E+00 5. 2.10E-02

Lungs 4.29E+00 .94E-01 4.52E-02

Skin 3.28E-03 1.27E+00

Spleen 3.28E-03 .15E-03 1.27E+00

Testes 2.47E-02 1.31E+00

Thymus 3.28E-03 .15E-03 1.27E+00

Thyroi 3.28E-03 1.27E+00

Uterus 3.28E-03 .15E-03 1.27E+00

Scenario 2 – Chronic inhalation of 15Bq/day of 239Pu Type S + acute intake of 10E+03 Bq: 

The second scenario modeled was a chronic inhalation of 15Bq/day of 239Pu Type S, 

MAD = 5 μm, through the 20 years of exposure with additional acute intakes of 

0,000 Bq/year, in the beginning of each year, of Type S 239Pu compound, AMAD = 1μm. The

ed

e

of

A

1

annual intake due to chronic exposure was 3.75 E+03 Bq, assuming that the worker was expos

250 days per year, and due to acute intake was 10E+03 Bq. If the procedure to estimate th

missed dose was applied, the calculated intake would be 2.72E+04 Bq for the whole period

employment, instead of real one that was 7.5E+04 Bq for chronic plus 2E+05 Bq for acute 

intakes. Since it is not significant chronic exposure, the worker could probably be monitored just 

once a year. Taking into account that the MDA value for the early 20 years, 0.57–0.51 dpm

(0.01 Bq), 239

Pu in urine would be detected just after 10 years of exposure, as seen in Table 2. 

The table also shows that whether the urine sample is collected in the end of the year the acute 

intake is not detected. As the contribution of chronic intake is not significant, the activities in 

urine due to acute intakes are very similar to the chronic intake of 15Bq/d. A graph below 

represents the data on Table 2. 

Table 2. Predicted Activity in Daily Urinary Excretion due to 20-Year Chronic 

Intake of 15 Bq/day Plus One Acute Intake of 10E+03 Bq Each Year 

239Pu in urinary excretion (Bq/d) Years Acute 10,000Bq/y Chronic 15 Bq/d 

1.00 6.43E-04 6.92E-04 1.

Chronic+acute 

33E-03 

2.00 1.26E-03 1.30E-03 2.55E-03

3.00 1.26E-03 2.07E-03 3.33E-03

4.00 1.83E-03 2.76E-03 4.59E-03

5.00 3.42E-03

3.34 E-03 7.

3.79E- E 8.42

2.87E-03

E-03

6.29E-03

6.00 35E-03 

7.00 

4.01

03 4.64 -03 E-03
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in u excreti

Table 2. Predicted Activity in Daily Urinary Excretion due to 20-Year Chronic 

Intake of 15 Bq/day Plus One Acute Intake of 10E+03 Bq Each Year 

239Pu rinary on (Bq/d) Yea Acute 10,0 C 1 Chron

03 5.15E 35E-03 

rs 00Bq/y hronic 5 Bq/d 

-03

ic+acute 

8.00 4.21E- 9.

9.00 03 5. E- E0 03 E- E0 03 2E- E0 03 7.13E- E0 03 7. E- E0 03 4E- E-02 

0 03 E- 51E-

.00 03 E- E17.00 03 9. E- E0 03 E- E-02 

0 03 E- E0 03 3E- E-02 

4.60E- 66

6.20

03 1.03 02 

10.0 4.99E- 03 1.12 02 

11.0

12.0

5.36E- 6.6 03 1.20 02 

5.72E- 03 1.28 02 

13.0

14.0

6.07E- 58

7.9

03 1.36 02 

6.41E- 03 1.43

15.0 6.74E7.06E8.37 03 1.

1.58

02 

16 8.78 03

03

02 

7.38E- 21

9.59

1.66 02 

18.0

19.0

7.69E- 03 1.73

8.00E8.30E9.95 03 1.79 02 

20.0 1.0 02 1.86

Chronic i 15 B gle s yea

00E

00E

1.00E

1.00E

1

rs

ntake of q/d + sin intake of 10000Bq once a r

1. -04

9

1.

i

-03

-02

-01

0.00 5.00 10.00 5.00 20.00 25.00

yea

P

u-2

3

n

urin

e (B

q/d)

acute 10000Bq/y

chronic 15Bq/d

Chonic+acute

cenario 3 – Chronic inhalation of 100 Bq/day of 239Pu Type S + acute intake of 10E+03 

Bq: The third scenario modeled was a chronic inhalation of 100 Bq/day of 239Pu Type S, 

AMAD = 5 μm, through the 20 years of exposure, with additional acute intakes of 10,000 

Bq/year, in the beginning of each year, of Type S 239Pu compound, AMAD = 1μm. The annual 

intake due to chronic exposure was 2.5E+04 Bq, assuming that the worker was exposed 250 days 

per year, and due to acute intake was 10+03 Bq, assuming that the worker was being monitored 

each 3 months. The calculated intake for the whole period of employment, was 5E+05 Bq for 

S
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chronic plus 2E+05 Bq for acute intakes. Table 3 shows that after the 2nd year of exposure, the 

urinary activity due to chronic intake is similar to the one in the first day after the acute intake of

10+03 Bq. It means that any acute intake in the order of 104 Bq may be confounded with chronic 

intake. In the 6th year of exposure, the contribution of chronic intake on the urinary activity is 

higher than the one due to acute intake of 10+03 Bq. As the urinary activity due to chronic 

intake is increasing over the years of exposure, it becomes difficult to detect acute intakes unless 

the measurement system is very sensitive or the intake is extremely high. The value of daily 

intake taken to simulate chronic exposure is just 25% higher than the one calculated based on the 

MDA value of 0.01 Bq (0.57dpm). In terms of dose, it is important for respiratory and 

gastrointestinal tract. A graph below represents the data on Table 3. 

Table 3. Predicted Activity in Daily Urinary Excretion due to 20-Year Chronic 

Intake of 100Bq/day Plus One Acute Intake of 10E+03 Bq Each Year 

239Pu in urinary excretion (Bq/d) 239Pu in urinary excretion (Bq/d) 

Years Acute 

10,000Bq 

Chronic 

100Bq/d Chronic+acute

Years Acute 

10,000Bq 

Chronic 

100Bq/d Chronic+acute

0.003 2.11E-02 2.02E-04 2.13E-02 8 4.38E-03 3.44E-02 3.88E-02 

0.25 8.92E-04 1.43E-03 2.32E-03 0.003 2.55E-02 3.44E-02 5.99E-02 

0.5 7.12E-04 2.48E-03 3.19E-03 0.5 4.98E-03 3.59E-02 4.09E-02 

02 

2

2

1.44E-02 0.003 2.70E-02 4.74E-02 7.44E-02 

3 1.83E-03 1.37E-02 1.55E-02 0.5 6.47E-03 4.85E-02 5.50E-02 

0.003 2.24E-02 1.37E-02 3.61E-02 13 6.28E-03 4.99E-02 5.62E-02 

0.25 2.13E-03 1.45E-02 1.66E-02 0.003 2.73E-02 4.99E-02 7.72E-02 

0.5 1.93E-03 1.58E-02 1.77E-02 0.5 6.79E-03 5.15E-02 5.83E-02 

0.75 1.86E-03 1.68E-02 1.87E-02 14 6.61E-03 5.27E-02 5.93E-02 

4 1.83E-03 1.80E-02 1.98E-02 0.003 2.76E-02 5.27E-02 8.03E-02 

0.003 2.35E-02 1.80E-02 4.15E-02 0.5 7.08E-03 5.42E-02 6.13E-02 

0.25 3.23E-03 1.93E-02 2.25E-02 15 6.91E-03 5.55E-02 6.24E-02 

0.5 3.01E-03 2.03E-02 2.33E-02 0.003 2.79E-02 5.55E-02 8.34E-02 

0.75 2.92E-03 2.13E-02 2.42E-02 0.5 7.38E-03 5.73E-02 6.47E-02 

5 2.87E-03 2.25E-02 2.54E-02 16 7.20E-03 5.85E-02 6.57E-02 

0.003 2.35E-02 2.25E-02 4.60E-02 0.003 2.82E-02 5.85E-02 8.67E-02 

25 3.23E-03 2.33E-02 2.65E-02 0.5 7.68E-03 5.95E-02 6.72E-02 

75 3.01E-03 2.45E-02 2.75E-02 17 7.50E-03 6.09E-02 6.84E-02 

6 2.92E-03 2.67E-02 2.96E-02 0.003 2.85E-02 6.09E-02 8.94E-02 

0.003 2.44E-02 2.67E-02 5.11E-02 0.5 7.97E-03 6.24E-02 7.04E-02 

0.5 4.04E-03 2.90E-02 3.30E-02 18 7.79E-03 6.38E-02 7.16E-02 

0.75 6.60E-04 3.36E-03 4.02E-03 9 4.83E-03 3.78E-02 4.26E-02 

1 6.43E-04 4.61E-03 5.25E-03 0.003 2.59E-02 3.78E-02 6.37E-02 

0.003 2.17E-02 4.61E-03 2.64E-02 0.5 5.38E-03 4.13E-02 4.67E-02 

0.25 1.53E-03 5.68E-03 7.21E-03 10 5.25E-03 4.12E-02 4.64E-02 

0.5 1.34E-03 6.67E-03 8.01E-03 0.003 2.63E-02 4.12E-02 6.75E-02 

0.75 1.28E-03 7.83E-03 9.11E-03 0.5 5.79E-03 4.29E-02 4.87E-02 

2 1.25E-03 8.96E-03 1.02E-02 11 5.60E-03 4.45E-02 5.01E-02 

0.003 2.24E-02 8.96E-03 3.13E-02 0.003 2.67E-02 4.45E-02 7.12E0.25 2.13E-03 1.02E-02 1.23E-02 0.5 6.16E-03 4.56E-02 5.18E-0

0.5 1.93E-03 1.13E-02 1.32E-02 12 5.96E-03 4.74E-02 5.34E-0

0.75 1.86E-03 1.25E-02
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Table 3. Predicted Activity in Daily Urinary Excretion due to 20-Year Chronic 

Pu retion (Bq/d) 

Intake of 100Bq/day Plus One Acute Intake of 10E+03 Bq Each Year 

239 in urinary excretion (Bq/d) 239Pu in urinary exc

Years Acute 

10,000Bq 100Bq/d Chronic+acute 0Bq/d Chronic+acute

7 3.92E-0 3.45E- 54E 02 

Chronic Years Acute Chronic 

10,000Bq 10

3 3.06E-02 02 0.003 2.88E-02 6. -02 9.42E0.003 2.50E-0 5.56E-02 03 6.89E- 02 

4.50E-03 3.71E-02 

2 3.06E-02 20 8.25E- 02 7.72E0.5 3.26E-02

Chronic intake of 100 Bq/d + single intake

10000Bq once a year

1.00E-04

1.00E-03

1.00E-02

1.00E-01

1.00E+00

4 5 6 7 8 9101 161718192021

yea s

 in

s of

0 1 2 3 112131415

P

u-2

3

8

urin

e (B

q/d)

acute 10000Bq

chronic 100Bq/d

Chronic

r

+acute

chronic inhalation of 100 Bq/day of 239Pu Type S, AMAD = 

μ through the 20 years of exposure, with additional acute intakes of 10,000 Bq/year, in the 

beg AD = 1μm. Sequential urinary 

me r ows that the values of 

rin y s factor of 1.5

Scenario 4 – Chronic inhalation of 100 Bq/day of 239Pu Type S + acute intake of 10E+03 Bq

- sequential measurements after the acute intake in the 10th year of exposure: 

The fourth scenario modeled was a 

5 m, 23 inning of each year, of Type S 9

Pu compound, AM

asu ements were performed one day after the acute intake. Table 4 sh

u ar excretion due to total intake are higher than the chronic ju t by a

(maximum). Depending on the recovery and measurement system, it may be part of the 

uncertainty of the measurement. 
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f

239Pu in urinary excretion (Bq/d) 

Table 4. Predicted Activity in Daily Urinary Excretion After an Acute Intake o

10E+03 Bq of Plutonium-239 in the Beginning of the 10th-Year Chronic Intake of

100 Bq/day Plus Annual Acute Intakes of 10E+03 Bq 

Days Acute 10,000 Bq/y Chronic 100 Bq/d Chronic+acute 

-730 2.55E-02 3.44E-02 5.99E-02 

-545 4.98E-03 3.59E-02 4.09E-02 

-400 4.83E-03 3.78E-02 4.26E-02 

-365 2.59E-02 3.78E-02 6.37E-02 

-180 5.38E-03 4.13E-02 4.67E-02 

1 2.63E-02 4.11E-02 6.74E-02 

2 3.70E-02 4.12E-02 7.82E-02 

3 2.97E-02 4.11E-02 7.08E-02 

4 2.23E-02 4.10E-02 6.33E-02 

5 1.71E-02 4.09E-02 5.80E-02 

6 1.37E-02 4.13E-02 5.50E-02 

10 8.27E-03 4.11E-02 4.94E-02 

20 6.62E-03 4.15E-02 4.81E-02 

30 6.49E-03 4.13E-02 4.78E-02 

50 6.34E-03 4.17E-02 4.80E-02 

90 6.15E-03 4.18E-02 4.79E-02 

100 6.11E-03 4.18E-02 4.79E-02 

200 6.05E-03 4.25E-02 4.85E-02 

250 5.95E-03 4.35E-02 4.94E-02 

300 5.85E-03 4.43E-02 5.01E-02 

365 5.80E-03 4.53E-02 5.11E-02 

Chronic intake of 100 Bq/d + single intakes of

10000Bq once a year - (10th year of exposure)

1.00E-03

1.00E-02

1.00E-01

1.00E+00

-500 -300 -100 100 300 500

days (10th year of exposure)

Pu-239 in urine (

B

q/d)

Acute

chronic

Chronic+acute
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CON

he incidental acute intake of high-fired may be difficult to identify 

• Low fraction of activity intake excreted through urine (10 Bq)

med 

s over the time of exposure, 

bviating the detection of incidental intakes, unless the activity is extremely high or the chronic 

exp u

 

These s

detecta of

high-fir

CLUSION: 

239 T Pu, in the first 20 years, 

because of several factors: 

• High MDA value (0.01Bq) 

• Urinalysis applied to evaluate incidental intake 

-6

According to workers’ interview, the urinalysis or other type of measurement was not perfor

following the accidents; in some cases after a considerable time after intake 

Fecal excretion associated with lung count is the best methodology to evaluate incidental intake,

but it was not performed. 

The contribution of chronic intake in urinary activity increase

o

os re is very low, or undetectable

cenarios show it would be unlikely that an acute intake of Type S 239Pu would be

ble, with the implication that it would be even more difficult to detect acute intakes

ed 239

Pu due to the lack of sensitivity of conventional urinalysis methods. 
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f

ssi m

in t w

produc

By virt

active e

reven re

rgely derived from a “learn by doing” process. Such a process, by its very nature would very 

like r

indicate

operati

• ere 

ing

ality limits and procedures for the safe handling of fissile 

materials, implementing the limits and procedures in all areas that handled fissile 

• 

laboratory. (Rothe 2005, pg. 45)

• on

er

periments conducted outside Building 886 were always subcritical.

(Rothe 2005, pg. 45)

• tu

same ones evaluating criticality safety throughout the plant. The 

n the

ATTACHMENT 10: LIKELIHOOD OF UNMEASURED NEUTRON 

EXPOSURES 

Rocky Flats was a “first of a kind” facility relative to large-scale processing and fabrication o

fi le aterials for nuclear weapons. For this reason, quantifiable knowledge of neutron hazards 

he orkplace at Rocky Flats was acquired, on an iterative basis concurrent with large-scale

tion and from the experience and experimental evidence from other sites. 

ue of the steep build-up and routine turnover of nuclear weapons and the constant and 

xperimentation with new weapons designs, the establishment of threshold values to 

p t criticalities and limit neutron exposures to workers during the 1950s and 1960s we

la

ly esult in missing neutron doses to a potentially significant number of workers. This is

d by recent history of the Rocky Flats criticality laboratory. From the inception of 

ons, Rocky Flats had a Nuclear Safety Group which: 

...was divided into two groups: the criticality mass laboratory, where experiments w

conducted, and criticality engineering. The principal functions of criticality engineer

included writing critic

materials, training and indoctrinating personnel who handled fissile materials, and 

performing auditing operations. (Rothe 2005, pg. 45) 

Until the early 1960s, criticality testing was done after-hours in the production glove

boxes. Experiments were only allowed to go towards criticality, but not allowed to go 

critical. Values were then extrapolated. The need to obtain more actual values was

recognized and in 1964, ground was broken on a state-of-the-art criticality mass

Investigators would set up the production materials in various arrays to perform neutr

multiplication experiments and make predictions with respect to safe geometries for 

various kinds of production vessels, spacing parameters, shipping containers, and oth

items. These in situ ex

Prior to constructing the [Critical Mass Laboratory in 1965], persons performing in si

experiments were the

Nuclear Safety Group was small; but everyone wore many hats. The division of 

manpower between experiments and plant safety only evolved during the late-1960s.

...At first, even this distinction was a little vague. Evaluators participated in a few 

experiments; and, to a lesser extent, experimenters became involved in plant-wide 

criticality safety. That distinction became much clearer during the early 1970s whe

AEC introduced the requirement that people performing critical experiments should

somehow be “certified.” (Rothe 2005, pg. 55) 
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Unlike ng the

period cky

Flats P tion of

binding ents. 

• For many years, the entire Nuclear Safety Group consisted of only 14 persons.

[emphasis added]This included three hired to perform experiments at the CML and one 

who served as a technical/mechanical/electronic support person. [the Nuclear Safety 

Group manager] had one secretary for the whole group and an Administrative Assistant. 

The remainder, always the larger group, worked closely with all buildings on plant site 

housing fissile materials to ensure nuclear criticality safety. (Rothe 2005, pg. 55)

• Early criticality safety advice did not possess the significance it now carries. 

Managers of various plant operations sought advice and guidance from [the Nuclear 

Safety Program director] and his team; but that input was neither considered binding 

nor limiting. Important safety information bore the innocent heading: “Nuclear Safety 

Recommendation.” That such important safety documents should ever have been 

relegated to the status of recommendation is somewhat surprising [Emphasis added]

(Rothe 2005, pg. 56)

• This author also recalls once finding an ancient and long-lost hidden sign suspended 

from some seldom-used equipment in a deep, never-visited, recess of one of 

 the plant’s more-remote production areas. This sign, discovered over 30 years ago, 

contained small hand-written black letters on a simple, white-painted, rectangle of

metal. The text was headed simply: ‘Crit Recommendation;’ and this was followed by 

a few terse words of advice. [Emphasis added] Had this been saved, it would have 

made an interesting comparison against modern postings. (Rothe 2005, pg. 56)

• As late as the mid-1960s, some limits were still issued verbally. Though not usually the 

case, last-minute changes in planned operations sometimes called for the Operations 

Manager to seek verbal modification of a written limit. This was occasionally given. 

Verbal criticality approvals were prohibited by later that decade. [Emphasis added.]

(Rothe 2005, pg. 58)

• ...the entire Nuclear Safety Group consisted of only 14 persons..... The division of 

manpower between experiments and plant safety only evolved during the late-1960s. 

(Rothe 2005, pg.55)

In addition to the inherent limitations of personnel monitoring for neutrons prior to the 

development of more optimal measuring technologies in the early 1970s, it is not at all clear that 

employees in process areas, particularly those whose jobs required moving in and around 

different processing areas (i.e., maintenance workers, pipefitters, sheetmetal workers, and 

electricians) were individually monitored for neutron exposure. The lack of individual 

monitoring for neutrons is inferred by the authors from a report summarizing data relative to the 

Former Radiation Worker Medical Surveillance Program at Rocky Flats in 2001: 

current criticality safety and neutron exposure protection regimes at DOE sites, duri

of peak production between 1955 and 1965, the system that was in place at the Ro

lant had a small staff and was not formalized relative to postings, the formaliza

operating limits in different process areas, and strict compliance with requirem
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Significant neutron radiation exposure was possible from alpha/neutron reactions 

with light elements, especially from plutonium tetrafluoride compound. Neutron 

exposure was possible from spontaneous fission of 240Pu during handling of 

large quantities of plutonium as metal, of 244Cm, or of plutonium enriched in 

240Pu for special projects conducted at the site. (Daugherty 2001)