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

July 2, 2007 

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 5: Transmittal of Draft SCA-SEC-TASK5-0056, 

Revision 1, Review of the Feed Material Production Center (FMPC) Special Exposure Cohort 

(SEC) Petition-00046 and the NIOSH SEC Petition Evaluation Report 

Dear Mr. Staudt: 

SC&A, Inc., is pleased to submit its draft report titled Review of the Feed Material Production 

Center (FMPC) Special Exposure Cohort (SEC) Petition-00046 and the NIOSH SEC Petition 

Evaluation Report, SCA-SEC-TASK5-0056, Revision 1. This revision of the draft report has 

been reviewed for Privacy Act information and edited accordingly. It is being presented to the 

Advisory Board for review. 

If you have any comment or questions, please contact me at 732-530-0104. 

Sincerely, 

John Mauro, PhD, CHP 

Project Manager 

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

Advisory Board Members H. Behling, PhD, SC&A 

L. Wade, PhD, NIOSH L. Elliott, NIOSH 

 H. Chmelynski, SC&A J. Neton, PhD, NIOSH 

 J. Fitzgerald, Saliant S. Hinnefeld, NIOSH 

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

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

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

L. Shields, NIOSH Project File (ANIOS/005/56) 

 
Draft 

ADVISORY BOARD ON 

RADIATION AND WORKER HEALTH 

National Institute for Occupational Safety and Health 

Review of the Feed Materials Production Center (FMPC) 

Special Exposure Cohort (SEC) Petition-00046 

and the NIOSH SEC Petition Evaluation Report 

Contract No. 200-2004-03805 

Task Order No. 5 

SCA-SEC-TASK5-0056, Rev. 1

Prepared by 

S. Cohen & Associates 

1608 Spring Hill Road, Suite 400 

Vienna, Virginia 22182 

Saliant, Inc. 

5579 Catholic Church Road 

Jefferson, Maryland 21755 

June 2007 

Disclaimer 

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

on Radiation and Worker Health (ABRWH) to maintain all possible openness in its 

deliberations. However, the ABRWH and its contractor, SC&A, caution the reader that at the 

time of its release, this report is pre-decisional and has not been reviewed by the Board for 

factual accuracy or applicability within the requirements of 42 CFR 82. This implies that once 

reviewed by the ABRWH, the Board’s position may differ from the report’s conclusions. Thus, 

the reader should be cautioned that this report is for information only and that premature 

interpretations regarding its conclusions are unwarranted.
Effective Date: 

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NOTICE: This document has been reviewed for Privacy Act information and has been edited accordingly. 

NOTICE: This report is pre-decisional and has not been reviewed by the Advisory Board for factual accuracy or 

applicability within the requirements of 42 CFR 82. 

Document No. 

SCA-SEC-TASK5-0056 

Effective Date: 

Draft — June 27, 2007 

S. COHEN & ASSOCIATES:

Technical Support for the Advisory Board on 

Radiation & Worker Health Review of 

NIOSH Dose Reconstruction Program Revision No. 

1 – PA Reviewed Draft 

Review of the Feed Materials Production Center 

(FMPC) Special Exposure Cohort (SEC) Petition00046 and the NIOSH SEC Petition Evaluation Report 

Page 2 of 142

Task Manager: 

________________________ Date: ___________ 

U. Hans Behling, PhD, MPH 

Project Manager: 

________________________ Date: ___________ 

John Mauro, PhD, CHP 

Supersedes: 

Rev. 0 – May 31, 2007 
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NOTICE: This report is pre-decisional and has not been reviewed by the Advisory Board for factual accuracy or 

applicability within the requirements of 42 CFR 82. 

TABLE OF CONTENT 

1.0 Introduction..........................................................................................................................5 

1.1 Scope and Purpose of SEC Review .........................................................................5 

1.2 Technical Approach and Review Criteria................................................................5 

1.3 Organization of the Report.......................................................................................7 

2.0 Key Information and Data Presented in the Fernald Site Profile.........................................8 

2.1 Feed Stocks and Materials Produced at FMPC........................................................8 

2.2 Production Facilities And Other Relevant Areas.....................................................9 

2.3 An Overview of NIOSH’s Approach to the Reconstruction of Internal 

Doses......................................................................................................................15 

2.3.1 Internal Doses Associated with Uranium Processes..................................15 

2.3.2 Internal Doses Associated With Thorium Processes.................................17 

2.3.3 Internal Doses Associated with K-65 Silo Processes ................................18 

2.4 External Doses from Beta, Gamma, and Neutron External...................................19 

3.0 Overview of the SEC Petition-00046 and NIOSH’s Evaluation Report ...........................20 

3.1 SEC Petition-00046 ...............................................................................................20 

3.2 SEC Petition Evaluation Report.............................................................................21 

3.3 NIOSH’s Response to Major Issues Raised in SEC-00046...................................22 

3.4 NIOSH Conclusions...............................................................................................24 

4.0 Preliminary Findings Associated with the Ability to Reconstruct Internal and 

External Doses for FMPC Workers...................................................................................25 

4.1 Findings Associated with Urinalysis Data for Assessing Intake of Uranium........25 

4.2 Findings Associated with Dose Assessments in Behalf of K-65 Wastes and 

Other Raffinate Wastes..........................................................................................36 

4.3 Findings Associated With Internal Dose Estimates for Thorium..........................49 

4.4 Generic Limitations and Findings Associated With the Use of Mivrlm Data 

for Uranium and Thorium Dose Estimates..........................................................103 

4.5 Findings Associated With External Exposure Monitoring At FMPC .................112 

5.0 Conclusions......................................................................................................................136 

6.0 References........................................................................................................................138 

Appendix A: Petitioner/Worker Interview..................................................................................142 

TABLES 

Table 2.2-1. Major Radionuclide Constituents in Pit Wastes................................................... 12 

Table 2.2-2. Isotopic Composition of K-65 Silos 1 and 2......................................................... 12 

Table 2.2-3. Radioactive Materials to Which Workers Were Exposed .................................... 14 

Table 4.1-1. Comparison of Total Fernald RU Contamination, TBD, Vol. 5 vs. DOE 2003... 36 

Table 4.2-1: Plant 6 Air Dust Data, 1960 to 1962..................................................................... 76 

Table 4.2-2: 50-Year Organ Dose Conversion Factors for U-234 and Th-232 – Inhalation .... 85 

Table 4.2-3. Number of In-vivo Measurements Performed Annually at the FMPC ............... 110 
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applicability within the requirements of 42 CFR 82. 

ATTACHMENTS 

Attachment 4.1-3............................................................................................................................29

Attachment 4.1-4A.........................................................................................................................31

Attachment 4.2-1A.........................................................................................................................40

Attachment 4.2-1B.........................................................................................................................41

Attachment 4.2-1C.........................................................................................................................42

Attachment 4.2-1D.........................................................................................................................43

Attachment 4.3-1A.........................................................................................................................55

Attachment 4.3-1B.........................................................................................................................56

Attachment 4.3-1C.........................................................................................................................57

Attachment 4.3-1D.........................................................................................................................58

Attachment 4.3-1E.........................................................................................................................59

Attachment 4.3-3A.........................................................................................................................62

Attachment 4.3-3B.........................................................................................................................63

Attachment 4.3-3C.........................................................................................................................64

Attachment 4.3-3D.........................................................................................................................66

Attachment 4.3-3E.........................................................................................................................67

Attachment 4.3-4A.........................................................................................................................71

Attachment 4.3-4B.........................................................................................................................72

Attachment 4.3-4C.........................................................................................................................73

Attachment 4.3-4D.........................................................................................................................75

Attachment 4.3-5A.........................................................................................................................81

Attachment 4.3-5B.........................................................................................................................82

Attachment 4.3-5C.........................................................................................................................83

Attachment 4.3-5D.........................................................................................................................84

Attachment 4.3-7............................................................................................................................87

Attachment 4.3-8............................................................................................................................90

Attachment 4.3-9A.........................................................................................................................93

Attachment 4.3-9B.........................................................................................................................95

Attachment 4.3-9C.........................................................................................................................96

Attachment 4.4-5..........................................................................................................................111

Attachment 4.5-1..........................................................................................................................113

Attachment 4.5-2A.......................................................................................................................120

Attachment 4.5-2B.......................................................................................................................121

Attachment 4.5-3A.......................................................................................................................123

Attachment 4.5-3B.......................................................................................................................124

Attachment 4.5-3C.......................................................................................................................126

Attachment 4.5-4A.......................................................................................................................129

Attachment 4.5-4B.......................................................................................................................132

Attachment 4.5-4C.......................................................................................................................133

Attachment 4.5-5..........................................................................................................................135
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applicability within the requirements of 42 CFR 82. 

1.0 INTRODUCTION 

1.1 SCOPE AND PURPOSE OF SEC REVIEW 

During the meeting of the Advisory Board (the Board) on Radiation and Worker Health held in 

Mason, Ohio, on February 8, 2007, S. Cohen & Associates (SC&A) was directed by the Board to 

perform a full review of the Feed Materials Production Center (FMPC) Special Exposure Cohort 

(SEC) Petition-00046 and the NIOSH SEC Petition Evaluation Report (ER) for said petition. 

Firstly, the scope of this review addresses specific issues of concern raised in the petition and 

NIOSH’s response to these concerns as given in the ER. (Because the ER makes frequent 

reference to (or defers to) the Fernald Site Profile (ORAUT-TKBS-0017), our review included 

the FMPC Site Profile.) 

Secondly, SC&A reviewed hundreds of documents that were considered relevant to the petition. 

documents reviewed include the following: 

• FMPC documents that were referenced and/or enclosed in the petition 

• Documents referenced/cited in the ER 

• Documents contained in the NIOSH Site Research Query Database 

The purpose of this review is to provide the Board with an independent assessment of issues and 

concerns that surround the petition and NIOSH’s response and proposed methods for 

accommodating these issues/concerns. Findings identified in our review are expected to provide 

the Board with a preliminary overview of potential issues that may impact the feasibility of 

dose assessment. Following a formal, multi-step resolution process, any unresolved findings 

may then be used by the Board for determining whether radiation doses can be estimated with 

sufficient accuracy, as defined in 42 CFR §83.13(c)(1): 

Radiation doses can be estimated with sufficient accuracy if NIOSH has 

established that it has access to sufficient information to estimate the maximum 

radiation dose, for every type of cancer for which radiation doses are 

reconstructed, that could have been incurred in plausible circumstances by any 

member of the class, or if NIOSH has established that it has access to sufficient 

information to estimate the radiation doses of members of the class more 

precisely than an estimate of the maximum radiation dose. 

1.2 TECHNICAL APPROACH AND REVIEW CRITERIA 

The approach used by SC&A to perform this review follows the protocols described in the draft 

report prepared by SC&A entitled Board Procedures for Review of Special Exposure Cohort 

Petitions and Petition Evaluation Reports, Revision 1 (SCA-TR-TASK5-0002, June 12, 2006) 

and the Report to the Working Group on Special Exposure Cohort Petition Review (Draft 

January 16, 2006). The latter is a set of draft guidelines prepared by a Board-designated working 

group for evaluation of SEC petitions performed by NIOSH and the Board. The former is a set 

of draft procedures prepared by SC&A and approved by the Board for use by SC&A on an 
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interim basis (Advisory Board meeting transcript of June 16, 2006, p. 132). The procedures are 

designed to help ensure compliance with Title 42, Part 83, of the Code of Federal Regulations

(42 CFR 83) and implement the guidelines provided in the report of the working group. 

Key review criteria identified in the report of the working group include the following: 

• Timeliness 

• Fairness 

• Understandability 

• Consistency 

• Credibility and validity of the data set, including pedigree of the data, methods used to 

acquire the data, relationship to other sources of information, and internal consistency 

• Representativeness and completeness of the exposure data with respect to the area of the 

facility, the time period of exposure, the types of workers and processes covered by the 

data 

The working group guidelines also recommend that NIOSH include in its SEC evaluation a 

demonstration that it is feasible to reconstruct individual doses for the cohort, including sample 

dose reconstructions. 

SC&A’s implementation of the SEC Review process includes the following steps: 

(1) Conduct a critical review of the petition and relevant reports, as well as documents and 

data that are enclosed and/or referenced in the petition/reports. For SEC Petition-00046, 

a major source of information included internal FMPC communiqués, letters, reviews, 

audits, and data sheets, which were cited in the petition as well as in the NIOSH Site 

Research Query Database. 

(2) Meet with and interview petitioner/former FMPC workers for the purpose of gaining 

additional insight into work practices, workplace conditions, monitoring methods, and 

dates of operations. 

(3) Identify additional issues/concerns that emerged from SC&A’s document review, which 

are independent of those stated in the petition. 

(4) As part of the SEC review, develop a preliminary technical position for issues identified 

in the petition, as well as SC&A’s independent findings. 

SC&A’s draft report with its preliminary findings will subsequently undergo a multi-step 

resolution process. Resolution includes a transparent review and discussion of draft findings 

with members of the Board’s working group, petitioner, claimants, and interested members of 

the public. This resolution process is intended to ensure that each finding is evaluated on its 
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technical basis in a fair and impartial basis. A final report will then be issued to the full Board 

for deliberation and a final recommendation. 

1.3 ORGANIZATION OF THE REPORT 

Following this introduction, Section 2.0 of this report provides summary data contained in the 

FMPC Site Profile. The site profile specifies relevant background information and methods to 

be used by NIOSH for the reconstruction of internal and external doses. Included herein are 

brief site profile summaries of materials and quantities processed, facility descriptions, and 

proposed methods for dose reconstruction. 

Section 3.0 of this report summarizes specific concerns and issues raised in the SEC Petition00046, as well as NIOSH’s Evaluation Report of the petition. In the Petition Evaluation Report, 

NIOSH provided responses to the petition’s concerns along with the conclusion that dose 

reconstruction is feasible for FMPC workers for the years 1951 through 1989. 

As a result of our review of the petition, NIOSH’s evaluation of the petition, the FMPC Site 

Profile, and other documents, SC&A identified a total of 29 findings, which are cited in Section 

4 of this report. In behalf of each finding, a discussion is provided that serves to explain the 

technical basis for our concern. For most findings, support is also provided by one or more 

FMPC documents, which are enclosed as attachment(s), or are referenced (see Reference List in 

Section 6.0). 

These attachments frequently contain empirical data and/or personal observations/opinions 

expressed by key individuals who were involved in FMPC operations, worker/workplace 

monitoring, and audits of the FMPC Health and Safety Program. As such, SC&A regards these 

historical documents as highly relevant, credible, and impartial. For this reason, the reader is 

encouraged to review the enclosed attachments and independently determine the degree to which 

they support each of the corresponding findings. For practical reasons, findings are grouped by 

category in the following subsection of Section 4.0: 

• Subsection 4.1: Findings associated with urinalysis data for assessing intake of uranium. 

• Subsection 4.2: Findings associated with dose assessments in behalf of raffinates and 

K-65 processes. 

• Subsection 4.3: Findings associated with internal thorium dose estimates. 

• Subsection 4.4: Limitations and findings associated with the use of MIVRML data for 

uranium and thorium dose estimates. 

• Subsection 4.5: Findings associated with external exposure monitoring at FMPC. 

Section 5.0 provides concluding comments regarding the impacts of our findings on dose 

reconstruction for FMPC workers.
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applicability within the requirements of 42 CFR 82. 

2.0 KEY INFORMATION AND DATA PRESENTED IN THE 

FERNALD SITE PROFILE 

The most current site profile for the Fernald Feed Materials Production Center (FMPC) consists 

of six Technical Basis Documents (TBDs) that were issued at various times (see Reference List 

for specific dates). These TBDs will be referenced throughout this report and include the 

following: 

• ORAUT-TKBS-0017-1: Introduction 

• ORAUT-TKBS-0017-2: Site Description 

• ORAUT-TKBS-0017-3: Occupational Medical Dose 

• ORAUT-TKBS-0017-4: Occupational Environmental Dose 

• ORAUT-TKBS-0017-5: Occupational Internal Dose 

• ORAUT-TKBS-0017-6: Occupational External Dose 

Collectively, the six TBDs of the site profile are intended to provide core information, data, and 

guidance that are intended to assist in the dose reconstruction of individual workers who may 

have been exposed to internal and external occupational radiation at FMPC, as stated in 

ORAUT-TKBS-0017-1: 

. . . This Profile can be used by dose reconstructors to evaluate internal and 

external dosimetry data for unmonitored and monitored workers and can serve as 

a supplement to, or substitute for, individual monitoring data. This document 

provides a site profile of FMPC that contains technical basis information to be 

used by the ORAU Team to evaluate the total occupational radiation dose for 

EEOICPA claimants. It provides information on buildings, operations, site 

conditions, modes and methods of potential radiological exposure, and inferred 

best estimates of dose parameters where data are missing or might be inaccurate.

[Emphasis added.] 

2.1 FEED STOCKS AND MATERIALS PRODUCED AT FMPC 

FMPC was a large-scale production facility that used a wide variety of complex chemical, 

metallurgical, and mechanical processes to convert various uranium and thorium feed stocks into 

products used by other facilities of the U.S. nuclear weapons complex. Operations began in 

1951, when the Atomic Energy Commission (AEC) contracted National Lead of Ohio (NLO) to 

process nuclear materials to produce high-purity uranium metal products in the form of derbies, 

ingots, billets, and fuel cores for other sites of the U.S. nuclear weapons complex. During its 38 

years of operation that ended in 1989, FMPC produced a total of 980,048 metric tons of uranium 

compounds such as uranium trioxide (UO3) and uranium tetrafluoride (UF4). Major feedstocks 

for uranium products included uranium ore (including pitchblende from the Belgian Congo), as 

well as scrap/recycled uranium (RU) shipped from other nuclear weapon sites. 

Although uranium processing was the principal function, FMPC also processed substantial 

quantities of thorium ore for the production of various thorium compounds and thorium metal. 

NIOSH estimated that a total of 2,855 metric tons of thorium were processed at Fernald but also 
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acknowledged that “. . . a large number of [thorium] records and files were destroyed in the early 

1970s during declassification efforts . . . data reconstruction indicates that thorium processing 

was limited to three plants over short periods in the 38-year production history of FEMP.” 

In addition to thorium that was processed at Fernald, the site was appointed as the national 

repository for thorium in 1972. NIOSH further assumed that “. . . Approximately, two-thirds of 

the material in the repository was processed at the FEMP . . . [and] the remainder originated at 

other DOE facilities.” 

According to the FMPC Site Profile, of the three FMPC plants that processed thorium, the 

largest quantity (71% or 2,030 MT) was processed in the Pilot Plant between 1964 through 1979. 

An estimated 456 MT were processed in Plant 9 in 1954 and 1955; and 369 MT were processed 

in Plant 8 in the years 1966, 1969, 1970, and 1971. 

2.2 PRODUCTION FACILITIES AND OTHER RELEVANT AREAS 

Of the 1,050 acres that collectively define the FMPC site, about 136 acres represent areas that 

include production facilities. These consist of the Pilot Plant and Plants 1 through 9. Their 

physical size, dates of operation, and principal functions/source materials are briefly described 

below. These parameters are highly relevant for assessing the adequacy of air sampling, which 

was the principal method employed by FMPC Health and Safety personnel for monitoring and 

controlling worker exposures. 

Pilot Plant. With a ground floor area of 23,500 ft2, this facility operated from 1951 through 

1989. Its principal function was to convert UF6 to UF4 for use in the uranium metal production. 

In addition to natural uranium, significant fractions of the UF6 feedstock were derived from 

depleted uranium generated at gaseous diffusion plants and enriched uranium associated with 

recycled uranium. During various periods of facility operation, processes shifted to the 

production of thorium compounds and thorium metal. 

Principal radioactive sources in the Pilot Plant were uranium in the form of UF4 that 

corresponded to natural, depleted, and enriched uranium, the radioactive daughters associated 

with pitchblende, and select radioactive contaminants contained in RU. 

Plant 1 – Sampling Plant. The Sampling Plant operated between 1953 and 1989 and has a 

ground floor area of 22, 040 ft2

. Principal functions include: (1) drying, crushing, milling, 

grinding and classifying feed materials for further processing, (2) sampling and storing large 

amounts of depleted, natural, and enriched uranium in open and covered storage areas, and (3) 

digesting enriched (5% to 20%) uranium -238 residues. Among the feed materials processed 

were Canadian ore containing thorium, pitchblende from the Belgian Congo (containing Ra-226 

and daughters), and recycled uranium containing Np-237, Pu-238/-239/-240/-241, Tc-99, Sr-90, 

and other contaminants. 

Plants 2/3 – Refinery. Built in 1953 and with a ground floor area of 36,604 ft2

, the refinery 

converted natural uranium ore and to a lesser extent, enriched recycled uranium to uranium oxide 

(UO3). The conversion of uranium produced a raffinate that contained the radioactive 
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contaminants associated with pitchblende and RU feed stocks. According to NIOSH’s SEC00046 ER, in 1968, Plant 2/3 was briefly used to produce thorium nitrate and thorium oxide. 

However, the ER admits that “. . . few details are available regarding the process . . .” 

Plant 4 – Green Salt Plant. This 26,500 ft2

 facility started operation in 1953 and operated until 

1989. Its primary function was to convert UO3 (produced in Plants 2/3) to green salt (UF4). This 

two-step process involved the reduction of UO3 to UO2 (or brown oxide) and the conversion of 

UO2 to UO4 with anhydrous hydrogen fluoride. Since the UO3 feed stock may have been 

derived from natural, depleted, and recycled uranium source material, the relative ratios of 

U-238, U-235, and U-234 in UO4 varied. 

Plant 5 – Metals Production Plant. At 58,620 ft2 of floor area, this facility operated between 

1953 and 1989. As part of the integrated production processes, Plant 5 received the UO4

produced in Plant 4 and converted the UF4 to pure uranium metal. The conversion of UF4 to 

depleted, natural, and enriched uranium metal derbies was achieved by a process developed at 

Ames Laboratory. The highly exothermic reduction of UF4 employed a magnesium chloride 

liner in electric resistance furnaces. Uranium derbies, in turn, had to be cleaned and recast into 

purified uranium ingots using vacuum induction furnaces. A fraction of ingots was also 

subjected to a variety of mechanical shaping by means of lathes, saws, grinders, and milling 

machines. 

Plant 6 – Metals Fabrication Plant. At 206,270 ft2 of floor area, the Metals Fabrication Plant was 

the largest. This facility operated between 1952 and 1989. Its primary function was to heat-treat 

uranium ingots and billet produced in Plant 5 to improve strength and grain structure. Heattreated uranium billets and ingots were also extruded, cut, and machined for various production 

reactors. 

(Note: As part of the site profile review, SC&A has identified the fact that pyrophoric stockpiles 

of thorium residues were oxidized in Plant 6 in order to “stabilize” this material. Oxidation of 

thorium residues was performed in the Plant 6 furnace from early 1960 to the middle of 1963.) 

Plant 7 – Hexafluoride Reduction Plant. Plant 7 began operations in 1954 but ceased operations 

in 1956. During the first 2 years, the principal function was to convert UF6 and UF4 (green salt) 

in a gas-gas reaction with hydrogen at 100°F. Between 1956 and 1969, the facility remained idle 

after which time it was used to store drums of green salt. 

Plant 8 – Scrap Metal Recovery Plant. This 25,500 ft2 facility operated between 1953 and 1989. 

In Plant 8, residues and scrap from processes involving enriched uranium were subjected to 

furnacing, which removed oil, graphite, water, and metallic impurities before being sent to the 

refinery for extraction and recovery of residual uranium. 

In addition to uranium residues, Plant 8 also processed thorium residues, which were first 

converted to thorium oxalate and then to thorium hydroxide. During the 4 years that included 

1966, 1969, 1970, and 1971, an estimated 369 MT of thorium hydroxide were produced. 

Plant 9 – Special Products Plant. This 48,500 ft2

 facility began operation in early 1954. For the 

first 2 years, the primary function of Plant 9 was to produce purified thorium metal by means of 
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processes developed in Iowa at the Ames Laboratory in which ThF4 was blended with calcium 

and zinc chloride in a bomb retort and heated. In the highly exothermic chemical reaction, ThF4

was reduced by the calcium to form a zinc-thorium derby. Following the de-zincing, the cleaned 

derby was remelted and recast into a purified thorium ingot. Ingots were further machined and 

shaped into a variety of thorium metal products. An estimated 380 metric tons of pure thorium 

metal were produced in the years 1954 and 1955 (and possibly into 1956). Thereafter operations 

at the Special Products Plant involved casting enriched uranium derbies and RU metal scrap into 

large diameter uranium ingots weighing up to 900 kg. Ingots were further subject to various 

shaping and machining processes. 

Support Facilities – Waste Management Facilities. Each of the 10 production facilities created 

liquid and solid wastes and/or scrap materials, some of which were recycled while others were 

stored/disposed onsite. Most relevant to potential internal and external exposures to workers 

include two concrete silos containing K-65 residues, six waste pits, one burn pit, and buildings 

64, 65, 67, and 68, which served as thorium storage facilities when FMPC was designated as 

DOE’s thorium repository. Summary descriptions of these facilities are provided below: 

• Thorium Buildings. Thorium compounds stored in Buildings 64, 65, 67, and 68 include 

thorium hydroxide, oxide, and oxalates that were generated in behalf of the aircraft 

nuclear propulsion (ANP) Program that was cancelled in 1961 and for the Light Water 

Breeder Reactor Program. Most of the stored thorium was in metal drums and cans that 

over time corroded, resulted in frequent fires, and required sampling and repacking. 

• Waste Pits. Waste Pits 1 through 6 were constructed and operated at various times during 

the 38 years of FMPC operations. Radionuclide contaminants in these pits varied 

significantly in terms of their relative, as well as absolute, concentrations and total 

amounts. For most waste pits, the isotopes of uranium (U-238, U-235, U-234) were the 

dominant contributors to activity levels. Pit wastes associated with the processing of RU 

feed stocks also contained variable amounts of transuranics, Tc-99, and some fission 

products. 

For waste pits 1, 2, and 4, radiothoriums contributed a large fraction to the total 

radioactivity; and for pits 3 and 5, radiothoriums dominated. Table 1 provides a 

summary of activities contributed by isotopes of uranium and thorium. 

Inspection of Table 2.2-1 shows that the vast majority of thorium activity in waste pits is 

contributed by Th-230, which is not linked to the production of thorium metal, thoria gel, 

and other thorium products, but is the radioactive daughter of U-234. Thorium-230 has a 

half-life of 80,000 years and like Th-232 and Th-228 is an alpha emitter. Sources of 

Th-230 in pit waste are waste streams from the processing of unrefined ores (e.g., 

Belgian Congo pitchblende and Canadian ores) that served as primary feedstocks for 

uranium products. 
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Table 2.2-1. Major Radionuclide Constituents in Pit Wastes 

Average Activity Concentration (pCi/g) of Waste Isotope Pit 1 Pit 2 Pit 3 Pit 4 Pit 5 Pit 6 

U-234 558 3867 208 783 636 3418 

U-235 56 1793 12 152 34 1042 

U-238 1951 4725 442 4644 641 16,975 

Th-228 20 63 12 115 31 <1 

Th-230 1504 1435 4638 428 4475 31 

Th-232 22 43 16 37 37 <1 

• Burn Pit. This facility served as a disposal site for pyrophoric materials between 1957 

and 1968. In 1984, contents of the burn pit were excavated and transferred to waste pit 4. 

Quantitative estimates of radionuclide-specific activity levels are not well documented, 

but may be assumed to parallel those of waste pit 4. 

• K-65 Silos. The processing of unrefined ores that included pitchblende from the Belgian 

Congo produced waste streams (or raffinates) that contained high levels of Ra-226 (and 

daughter products) as well as Th-230. Large concrete tanks measuring 27 feet high and 

88 feet in diameter were constructed in 1951 and 1952 for the interim storage of raffinate 

wastes. Pitchblende raffinates generated in Plant 2/3 were disposed by direct pipelines 

into Silos 1 and 2. 

In addition to raffinates generated at FEMP, African pitchlende had also been processed 

in earlier years by Mallinckrodt Chemical Works (MCW). Due to a shortage of storage 

space, raffinates from MCW were shipped in 55-gallon drums in 1951 and 1952 for 

interim storage. With Silos 1 and 2 completed in July of 1952, the contents of 

approximately 13,000 55-gallon drums from MCW were transferred into Silos 1 and 2 

over a 6-year period that started in July 1952 and ended in September 1958. 

The transfer of 13,000 drums of MCW raffinates into the K-65 Silos 1 and 2 was a 

process performed by laborers, who manually transferred the drums’ contents onto a 

conveyor belt that dropped raffinate waste into the storage silos. The total amount of 

raffinates stored in Silos 1 and 2 is estimated at 10,000 metric tons. In 1993, core 

samples taken from Silos 1 and 2 yielded activity values summarized in Table 2.2-2. 

Table 2.2-2. Isotopic Composition of K-65 Silos 1 and 2 

Silo 1 Silo 2 Isotope Activity (nCi/gm) Activity Fraction Activity (nCi/mg) Activity Fraction 

Uranium – Total 1.68 1.61 E-3 2.37 3.04 E-3 

Ac-227 (β) 7.67 7.36 E-3 6.64 8.50 E-3 

Pa-231 (α) 4.04 5.17E-3 

Pb-210 (β) 202 1.94 E-1 190 2.43 E-1 

Po-210 (α) 281 2.70 E-1 231 2.96 E-1 

Ra-226 (α) 477 4.58 E-1 263 3.36 E-1 

Th-228 (α) 2.28 2.19 E-3 7.36 9.42 E-3 

Th-230 (α) 68.9 6.62 E-2 76.2 9.75 E-2 

Th-232 (α) 1.11 1.07 E-3 0.99 1.26 E-3 
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From data shown in Table 2.2-2, the following summary conclusions may be drawn: 

• Ra-226 and the longer-lived daughters Pb-210 and Po-210 contribute the majority of 

activity. 

• The specific activity in Silo 1 is significantly higher than that of Silo 2. 

• There is a clear disequilibrium condition between Ra-226 and Po-210/Pb-210, which 

suggests that about 50% of Rn-222 is/has been released from the raffinate waste. 

• As was the case for waste pits, thorium activity is dominated by Th-230. 

In summary, FMPC was a large-scale integrated processing facility, which utilized various feed 

stocks of uranium and thorium. In turn, these processes produced a complex variety of 

intermediate products, waste streams, and finished products, which differed in their radiological, 

chemical, and physical properties. Table 2.2-3 identifies the various radiological source 

materials to which an estimated 7,000 FMPC workers were exposed. 
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Table 2.2-3. Radioactive Materials to Which Workers Were Exposed 

Uranium – U-238, U-235, U-234, U-236 

 - Uranium ores from mines and mills 

 - Pitchlende ore from Belgian Congo 

 - Recycled uranium from spent fuel elements and transuranic production targets 

Radioactive Daughters of Uranium Ores/Pitchblendes: 

 - Th-230 

 - Ra-226 

 - Rn-222 

 - Pa-231 

 - Ac-227 

 - Rn-220 

Contaminants Associated with RU: 

 - Pu-238, Pu-239, Pu-240, Pu-241 

 - Np-237 

 - Tc-99 

 - Sr-90 

Enrichments of Uranium: 

 - DU 

 - Unat

 - EU (up to 20%) 

Chemical Forms of Uranium: 

 - Uranium metal 

 - UO2 – uranium-oxide high-fired; brown oxide 

 - U3O8 – uranium-oxide/yellowcake 

 - UO4 – uranium tetraoxide 

 - UF4 – green salt 

 - UF6 – uranium hexafluoride gas 

 - UO3 – uranium trioxide (orange oxide) 

 - UO2F2 – uranyl fluoride 

 - UO2(NO3)2 – uranyl nitrate 

Thorium and Daughters: 

 - Th-232 and Th-228 

 - Ra-228 

 - Ac-228 

 - Pb-212 

 - Rn-220, etc. 

Thorium Chemical Forms: 

 - Th-metal 

 - ThO2 – oxide 

 - Th(OH)4 – hydroxide or gel 

 - ThNO3 – oxalate 
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2.3 AN OVERVIEW OF NIOSH’S APPROACH TO THE RECONSTRUCTION OF 

INTERNAL DOSES 

 

2.3.1 Internal Doses Associated with Uranium Processes 

ORAUT-TKBS-0017-5 of the Fernald Site Profile provides the technical basis for estimating 

occupational internal dose to FMPC workers associated with uranium and thorium processes. 

This Technical Basis Document (TBD) contains pertinent information and assumptions for use 

by dose reconstructors. Due to the focused scope of an SEC evaluation, a comprehensive review 

of this TBD is neither warranted nor essential. Our review is, therefore, limited to the 

identification of core information, methods, and assumptions stated in the TBD that affect the 

ability to reconstruct internal doses. For simplification, direct quotations from the TBD Section 

5 are cited when appropriate: 

 From Sections 5.1 and 5.2:

(1) The original health and safety program was conducted with an industrial 

hygiene emphasis, based upon uranium heavy metal toxicology . . . Basic 

changes in the radiological protection program occurred with the contractor 

change in 1986 with the addition of radiation safety staff and a greater emphasis 

on radiation protection principles. [Emphasis added.] 

(2) Due to the inherent nature of the processes, the limitations of the ventilation and 

material confinement systems, and the volume (and mass) of the materials, 

significant environmental and in-plant releases of radioactive materials 

occurred during FEMP operations. The work environs included a 

continuous/chronic potential for internal exposure, as demonstrated by the 

comprehensive air monitoring program and the urine sampling program for 

uranium (documented by air sample and urine uranium data sheets). [Emphasis 

added.] 

(3) There are approximately seven steps in the process of conversion of uranium ore 

or other scrap recovery materials to metallic uranium. Those steps produce a 

number of compounds, each of which has specific chemical characteristics that 

are associated with different internal exposure parameters. Each of the 

compounds identified in Table 5-4 was handled in MT quantities. Most of the 

compounds were dry powder or granular in form and represented a dust hazard

potential as the material was processed, transferred, and otherwise handled.

[Emphasis added.] 

(4) Production operations that involved handling dry uranium materials were 

generally equipped with engineered ventilation systems for controlling dusts. 

Standard operating procedures required the use of respiratory equipment when 

dusty conditions were anticipated. Good housekeeping involving the immediate 

cleanup of spilled uranium products was also a standing policy and practice. In 

spite of this emphasis on engineered and administrative contamination controls 

and policy to reduce the release of radioactive materials, spills and routine 
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releases occurred. In addition to the routine releases at FEMP, there were 

frequent “upset” conditions (i.e., spills, effluent filter ruptures, etc.) that 

produced episodic airborne radioactivity in the work areas and plant effluents, 

and were of a magnitude that the ventilation systems were unable to contain all 

of the releases. [Emphasis added.] 

(5) A radiological controls program was in place from the beginning of FEMP 

operations. The internal dose control program consisted of: 

• An air sampling program in all processing areas to evaluate internal 

exposure potential via inhalation 

• Urine samples submitted after at least a two-day work break to allow 

elimination of uranium cleared rapidly via the GI tract (this material 

causes relatively little dose) 

• In vivo analysis once a month for high exposure-potential workers on a 

frequent urinalysis program and once a year for workers with a low 

potential for internal intake.

(6) [For] uranium . . . its chemical toxicity can be the dominant hazard in cases of 

readily or moderately soluble compounds of depleted, natural, and low-enriched 

uranium. . . 

The early basis for conducting routine urine analysis was to assure that uranium 

exposure controls were adequate to prevent chemical toxicity. . . . 

Though the primary exposure control was based on toxicity concerns, some 

radiological exposures were reported in addition to a few work restrictions based 

upon radiological dose limits. The radiological dose determinations were based 

upon in vivo lung counting. In vivo lung counts were routine from 1968 to 1989 

using a Mobile In Vivo Radiation Monitoring Laboratory (MIVRML) from 

ORNL Y-12. . . . The results were reported in milligrams of 235U and total 

milligrams of uranium (mg U) . . . [Emphasis added.] 

In summary, these statements imply the following: 

• Although general air (GA) and breathing zone (BZ) air sampling was regarded by FMPC 

personnel as the primary means for controlling intakes below levels that could result in 

chemical toxicity to the kidneys, NIOSH has stated that, for dose reconstruction, 

photofluorometric urine bioassay data will be used as the primary tool for estimating 

internal exposure to uranium. In lieu of or in support of urinalysis data, in vivo lung 

counts and air sampling may also be used for dose reconstruction. 

• Only uranium urinalysis was performed routinely from the 1950s to 1986. Thus, there 

were no direct measurements of either TRU contaminants or radioactive uranium 
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daughters, e.g., Th-230, Ra-226. (SC&A notes that a limited number of radon breath 

measurements were taken.) 

• Because urinalysis was based on a photofluorometric method and reported in units of mg 

U/liter urine, the isotopic mixture of uranium isotopes is unknown. 

• To account for RU contaminants (that include plutonium out of specification P005), the 

TBD allows for the presence of Pu-239, Np-237, and Tc-99 in concentrations of 100, 

3,500, and 9,000 ppb, respectively, as given in Table 5-12 of the TBD. 

• Before DOE Order 5480.11 (effective in 1989), bioassay data at Fernald were not used to 

estimate intakes or internal organ doses. 

• In 1968, in vivo lung monitoring began with the ORNL Y-12 MIVRML, which 

continued until 1989. Results of in vivo lung counting were calibrated in μCi of U-235 

but reported in mg of U in the lung. The conversion of μCi to mg total uranium assumed 

a 1% enrichment. 

2.3.2 Internal Doses Associated With Thorium Processes 

Section 5.2.3 of the TBD acknowledged that: 

Much of the thorium production data has been lost, and the plant and bioassay

monitoring data recovered to date has been sparse. . . . [and] that a large number 

of records and files were destroyed in the early 1970s during declassification 

efforts . . . 

A fundamental difficulty of dose reconstruction for thorium processing is that 

either 1) in vitro bioassays for thorium were not performed or 2) data is not 

available until after 1986. An additional consideration is that air sampling data 

was not used to calculate intake and dose until after 1986. Air monitoring was 

used only to control exposures to levels below the MAC. [Emphasis added.] 

Given these limitations, the TBD recommends a “claimant-favorable default exposure” approach 

for assigning thorium intakes. This default approach model assumes the following: 

• An intake exposure period of 100 hours per year at an air concentration of 10 MAC 

• An intake for an exposure period of 500 hours per year at an air concentration of 

0.1 MAC 

• No respiratory protection 

Using these assumptions, the default thorium intake model assigns a total of 1050 MAC-hours 

per year for thorium-232 and thorium-228, which are assumed to exist in secular equilibrium. 

This translates to an annual inhalation intake of about 30 nCi each for Th-232 and Th-228. 
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The TBD, however, states that this claimant-favorable default model may not be used for 

workers with in vivo chest counting data. The following guidance is provided: 

. . . When available, the chest count data should be used to constrain an 

employee’s intake. In other words if there are chest count results for an employee, 

the smaller of the default thorium intake or the chest count determined intake 

should be assigned for full dose reconstructions. [Emphasis added.] 

SC&A interprets these recommendations to imply that when empirical in vivo monitoring data 

suggest an intake that exceeds the 1050 MAC-hour default value, the dose reconstructor should 

ignore monitoring data and assign the smaller default-modeled dose of 1050 MAC-hours for 

dose reconstructions involving best estimates. 

2.3.3 Internal Doses Associated with K-65 Silo Processes 

From 1953 until 1958, pitchblende ore containing high concentrations of uranium and uranium 

daughter products was used as the raw material for the production of uranium products. This ore 

was obtained from the Shinkolobwe mine in the Belgium Congo. Uranium was separated from 

the pitchblende ore by the use of the three-phase Purex process: digestion, extraction, and 

denitration. The aqueous raffinate or waste from this process was pumped into one of two large 

concrete silos for storage. These residues, which were assigned the code name “K-65,” contain 

small amounts of Th-232 and Th-228 and daughter products of uranium including Th-230, 

Ra-226, Rn-222, Pb-210, and Po-210. The raffinate was slurried from the refinery (Plant 2/3) 

through pipes into Silo 2. 

The radium containing material in K-65 Silo 1 came from another source. Prior to 1952, large 

amounts of radium bearing radioactive waste were shipped to the Fernald Site from Mallinckrodt 

Chemical Works in St. Louis, and eventually stored in the K-65 Silo 1. When the waste material 

first arrived, however, it was placed in metal drums, which were temporarily stored on a concrete 

pad near Plant 1. An internal FMPC memorandum indicated that 13,000 55-gallon drums of 

K-65 material were received at the FMPC in the period September 25, 1951 to July 31, 1952. 

Other drums of K-65 material were stored in Plant 8 for long periods of time. The drummed 

material was transferred into Silos 1 and 2 over a 6-year period between July 1952 and 

September 1958. 

The transfer of K-65 waste material from the 13,000 drums was a manual process that exposed 

workers internally to high airborne levels of contaminants, as well as externally to penetrating 

radiation. 

Due to insufficient internal monitoring data associated with the K-65 processes, the TBD 

provides a default dose model that is described as “claimant-favorable and bounding.” In 

addition to the default particulate intake model, the TBD also provides a means for modeling 

radon/radon daughter exposures associated with the initial lid removal of the 13,000 drums. Key 

parameters of the model assume a Rn-222 air concentration of 230 pCi/l in full equilibrium with 

the short-lived daughter products for an annual exposure of 2.9 WLM. 
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For “environmental” releases of radon from the K-65 silos, additional radon exposures to 

workers are based on an atmospheric dispersion model that assumes an annual release of about 

5,000 Ci of radon from Silos 1 and 2. 

2.4 EXTERNAL DOSES FROM BETA, GAMMA, AND NEUTRON EXTERNAL 

For external penetrating radiation to the whole-body and for extremity/skin exposures, NIOSH 

intends to assign doses that reflect empirical measurements of personnel dosimeters that include 

film and thermoluminescent dosimeters. 

Because neutron exposures were not monitored at FMPC, potential neutron doses will be based 

on a neutron-to-photon ratio model, which has as its 95th percentile value, a neutron-to-photon 

ratio of 0.23.
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3.0 OVERVIEW OF THE SEC PETITION-00046 AND NIOSH’S 

EVALUATION REPORT 

3.1 SEC PETITION-00046 

SEC Petition-00046 qualified on April 6, 2006. The petition requested that NIOSH consider the 

following class: “All employees and all sub-contractors who worked at all locations at the Feed 

Materials Production Center (FMPC) in Fernald, Ohio, also known as the Fernald Environmental 

Management Project (FEMP), during the time period of January 1, 1951 through December 31, 

1989. 

While the core of the petition is relatively brief, the petition makes reference to and includes a 

large number of documents that support the petition’s concerns. Critical elements defined in 

SEC-00046 include the following concerns: 

• Issue #1: Failure to Monitor Internal Exposure to Contaminants in Recycled Uranium. 

Important contaminants for which there are no monitoring data include Pu-239/-240, Np237, and Tc-99. 

• Issue #2: Insufficient Monitoring of Worker Exposures to Thorium. Concerns raised in 

the petition focused on the absence of urinalysis monitoring for thorium, the incomplete 

documentation regarding the locations and time periods of thorium processing at FMPC, 

and thus, the incomplete air monitoring for all locations and all time periods. 

• Issue #3: Failure to Monitor Internal Exposure to Radioactive Daughter Products 

Contained in Raffinates of Uranium Extracted from African Pitchblende and Canadian 

Ores. Identified among the radionuclides of concern were Ra-226 and Rn-222, along 

with its short-lived and long-lived daughter products. 

• Issue #4: Failure to Monitor Exposure to Neutrons. Given the large quantities of alphaemitting radionuclides processed at FMPC, concerns were raised about unmonitored 

neutron exposures resulting from alpha, neutron reactions between isotopes of U/Th, and 

atoms with low atomic number (e.g., fluorine, beryllium). 

• Issue #5: Inappropriate Assumption Regarding the Protective Role of Respiratory Use 

During K-65 Silo Processes. 

• Issue #6: The Historical Practice by FMPC to Control/Limit Worker Exposures Only to 

Levels Below Those Associated with Chemical Toxicity and With No Attempt to 

Convert Urinalysis and Air Sampling Data to Dose Estimates. 

• Issue #7: Failure to Include Fecal Analysis as Part of a Routine Bioassay Program for 

Monitoring FMPC Workers. 

• Issue #8: Falsification/Manipulation of Workplace Air Sampling. 
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3.2 SEC PETITION EVALUATION REPORT 

On October 25, 2006, NIOSH issued its final SEC Petition Evaluation Report (ER) for SEC00046. As stated in Section 1.0 of the ER, this report evaluated “. . . the feasibility of 

reconstructing doses for all subcontractors who worked at all locations at the Feed Materials 

Production Center (FMPC) in Fernald, Ohio . . .” [Emphasis added.] 

Under 42 CFR § 83.13(c)(1), the feasibility to reconstruct doses includes: 

. . . radiation doses [that] can be estimated with sufficient accuracy if NIOSH has 

established that it has access to sufficient information to estimate maximum 

radiation dose for every type of cancer for which radiation doses are 

reconstructed, that could have been incurred in plausible circumstances by any 

member of the class, or if NIOSH has established that it has access to sufficient 

information to estimate the radiation dose of members of the class more precisely 

than an estimate of the maximum radiation dose. [Emphasis added.] 

The Evaluation Report responded to specific concerns and issues raised in the SEC Petition00046 as summarized in Section 3.1 above and concluded that: 

. . . NIOSH has established that it has access to sufficient information to: (1) 

estimate the maximum radiation dose incurred by any member of the class; or (2) 

estimate radiation doses more precisely than a maximum dose estimate. 

Information available from the site profile and additional resources is sufficient 

to document or estimate the maximum internal and external potential exposure to 

members of the proposed class under plausible circumstances during the specified 

period. [Emphasis added.] 

Furthermore, 

Per EEOICPA and 42 C.F.R. § 83.13(c)(3), NIOSH need not make a health 

endangerment determination, as it has determined that it has sufficient 

information to estimate dose for the members of the proposed class.

Data and information employed by NIOSH in its evaluation are cited in Section 4.0 of the ER. 

The recommendation to deny the SEC status was principally based on information provided in 

Technical Basis Documents ORAUT-TKBS-0017-1, -2, -3, -4, -5 and -6, which collectively 

define the FMPC Site Profile. As previously quoted in Section 2.0 above, NIOSH stated that: 

Dose reconstructors can use this Site Profile to evaluate internal and external 

dosimetry data for monitored and unmonitored workers, and to supplement or 

substitute for, individual monitoring data. [Emphasis added.] 

Other supportive documents reviewed and used by NIOSH to support its recommendation to 

deny the SEC status include: 
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• ORAUT-OTIB-0006, Rev. 03; ORAUT-OTIB-0024, Rev. 00; ORAUT-PROC-0060, 

Rev. 01; and ORAUT-PROC-OO61, Rev. 01 

• A total of 1,628 documents contained in the NIOSH Site Research Database as pertaining 

to the FMPC site 

• Fernald Historical Records 

• Documentation and/or Affidavits provided by petitioners 

Sections 5.0 and 6.0 of the ER provide summary descriptions of FMPC processes, FMPC 

monitoring practices, and available monitoring data. These data closely parallel information 

contained in the six TBDs that define the FMPC Site Profile and provide the technical basis for 

Section 7.0 of the ER. 

While Sections 7.1, 7.2, and 7.3 of the ER address the generic feasibility of internal and external 

dose reconstruction for FMPC workers, Section 7.4 addresses specific issues and concerns 

identified in the SEC-00046 petition, as summarized below. 

3.3 NIOSH’S RESPONSE TO MAJOR ISSUES RAISED IN SEC-00046 

Response to Issue #1

NIOSH admits that beginning in 1961, recycled uranium was introduced at FMPC as feedstock 

(inclusive of POOS); however, no analyses for the presence of RU contaminants were conducted 

in behalf of urine or air sample analysis. 

In an attempt to address these monitoring deficiencies, the ER merely states that: 

Section 5.2.2 of Technical Basis Document for the Fernald Environmental 

Management Project (FEMP)-Occupational Internal Dose provides an approach 

to account for missed internal dose from unmonitored or undetected recycled 

uranium impurity activities. This approach determines the uranium intake, and 

then (for intakes occurring after 1961) adds a claimant-favorable ratio of 

recycled uranium impurity activities to that intake.

Response to Issue #2

NIOSH’s position is that for any individual who had the potential for thorium exposure, the 

default intakes will be assigned as described in Section 5.2.3 of the TBD. The default 

assumption assigns 1,050 MAC-hours per year for a radionuclide mixture defined in Table 5-16. 

(Note: The TBD contains two tables designated at 5-16; one on page 24 and the other on 

page 25.) 
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However, the default assignment of 1,050 MAC-hours per year will likely be applied only to 

maximized (i.e., non-compensable) dose reconstructions. The ER states that “. . . in light of 

new data, a more precise approach is being developed.” [Emphasis added.] 

NIOSH’s more precise approach will assumedly apply to dose reconstructions that are based 

on best estimates and employ post-1968 MIVRML lung counting data and/or air monitoring 

data. 

Response to Issue #3

NIOSH intends to bound worker exposures to Ra-226, Rn-222, and other radioactive daughter 

products contained in pitchblende ores by a maximizing approach that is applicable to K-65 silo 

operations. This maximizing approach is described on pages 24 through 28 of the TBD 

(ORAUT-TKBS-0017-5) and combines a limited amount of empirical data with various 

assumptions. 

For non-K-65 Silo workers, other data may be used that includes radon breath analyses and the 

results of a FMPC radon study published in 2004. 

Response to Issue #4

For assigning missed neutron doses to select worker groups, NIOSH intends to employ a 

neutron-to-photon ratio of 0.23, as described in Section 6.3.5.2 of the TBD. This value is 

described as the 95th percentile for a lognormal distribution involving paired sets of photon and 

neutron measurements on 56 individual drums of UF4. 

Response to Issue #5

For assigning estimates of internal exposures to K-65 workers, the ER states that “. . . the 

maximizing approach to bounding K-65 silo exposure does not take credit for respiratory 

protection (ORAUT-TKBS-0017-5).” [Emphasis added.] 

Response to Issue #6

In spite of the fact that pre-1989 urinalysis employed a fluorophotometric method that only 

defined the elemental uranium content in urine in units of mg/liter, these data can be readily 

converted to activity values in behalf of depleted, natural, and low-enriched uranium, as dictated 

by worker location/work processes. Once converted to radiological units, both urine and air 

sampling data can be used to derive internal doses. 

Response to Issue #7

The ER states that while fecal analysis may have been employed at FMPC as part of select 

incident investigations, the decision to exclude fecal analysis from routine use was based on the 

limited reliability of this bioassay technique.
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Response to Issue #8

Documents contained in the petition (in the form of an affidavit and a report) imply that air 

sampling and effluent releases may have been manipulated and/or misrepresented. Upon review 

of these documents, NIOSH has concluded that: 

. . . this practice was unlikely to have routinely occurred, and since NIOSH will 

not be relying on a single air sample result to estimate a worker’s intake (but 

rather a distribution of or compilation of multiple air dust measurements), it is 

unlikely that this practice would have a significant affect on an individual’s dose. 

[Emphasis added.] 

3.4 NIOSH CONCLUSIONS 

The ER concluded that based on available information that includes monitoring records, process 

descriptions, and source term data, dose reconstruction is feasible for FMPC workers employed 

between January 1951 and December 1989 who may have been exposed internally and 

externally to the following sources: 

• Internal

- Uranium

- Thorium

- Other radionuclides (e.g., POOS nuclides, radon, thoron) 

• External

- Gamma 

- Beta 

- Neutron 

- Occupational medical 

On a final note (and perhaps to support its conclusions), the ER states the following: 

As of September 14, 2006, a total of 690 claims have been submitted to NIOSH 

for individuals who worked at FMPC during the years identified in the proposed 

class definition. Dose reconstructions have been completed for 619 individuals 

(90%).
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4.0 PRELIMINARY FINDINGS ASSOCIATED WITH THE ABILITY 

TO RECONSTRUCT INTERNAL AND EXTERNAL DOSES 

FOR FMPC WORKERS 

 

This section of the SEC review process identifies discrete issues of concern that may adversely 

affect the ability to estimate FMPC worker exposures from internal and external sources. 

Findings presented below are grouped as follows: 

• Section 4.1: Findings Associated with Urinalysis Data for Assessing Intake of Uranium 

• Section 4.2: Findings Associated with Dose Assessments in Behalf of K-65 Wastes and 

Other Raffinate Wastes

• Section 4.3: Findings Associated with Approaches for Internal Thorium Dose Estimate 

• Section 4.4: Limitations and Findings Associated with the Use of MIVRLM Data for 

Uranium and Thorium Dose Estimates 

• Section 4.5: Findings Associated with External Exposure Monitoring at FMPC 

4.1 FINDINGS ASSOCIATED WITH URINALYSIS DATA FOR ASSESSING 

INTAKE OF URANIUM 

Finding 4.1-1: Limitations Associated with the Use of Fluorophotometric Urinalysis Data 

Section 5.3.6 of FMPC TBD (as well as Section 6.1 of the SEC-00046 Evaluation Report) states 

the following: 

. . the fundamental and primary bioassay for the first 35 years (1951 – 1986) of 

Fernald operational experience was urine analysis for uranium metal, reported 

in milligrams per liter. [Emphasis added.] 

Thus, it is NIOSH’s intention to rely primarily on urinalysis data of elemental uranium to 

assess organ doses that may have resulted from the internalization of U-238, U-235, and U-234 

of variable degrees of enrichment, which, moreover, existed in various chemical forms that 

include UF4, UF6, UO2F2, UO2 (NO3)2, UO3, UO2, UO4, U3O6, and U metal. 

By design, the fluorophotometric urinalysis merely measures the total amount of elemental 

uranium without regard to the isotopic composition and radioactivity levels of the excreted 

uranium. Justification for this simplistic monitoring approach of FMPC workers was based 

solely on concerns for the chemical toxicity of uranium on kidney function, as opposed to 

radiological concerns. 

Thus, the obvious difficulty with the conversion of urine data to organ-specific radiation dose(s) 

is the near absence of critical data. Unknown data include the relative isotopic composition(s) of 

U-238/U-235/U-234 of materials representing depleted, natural, recycled/enriched uranium as 
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well as the chemical form and solubility of uranium materials. Over even a modest employment 

period, select workers such as members of a labor pool or roving maintenance personnel must 

reasonably be expected to have been exposed to a wide variety of uranium materials. 

Neither the FMPC TBD nor the SEC-000046 ER addresses the complexity of this problem or 

provides specific guidance for overcoming these deficiencies in the reconstruction of “best 

estimate” doses. 

Finding 4.1-2: The Questionable Integrity of Fluorophotometirc Urinalysis Data 

Aside from the intrinsic limitations that characterize the fluorophotometric urinalysis method 

discussed in Finding 4.1-1, there are reasons to also question the integrity/accuracy of the 

reported results. In addition to the near absence of formal records that define the analytical 

protocols, quality assurance/quality controls, instrument(s) performance standards (e.g., limits of 

detection), training and qualifications of laboratory personnel, etc., there is reason for concern 

about the integrity of reported results that reflect the perceived role of the urinalysis program by 

the Health and Safety personnel at FMPC. 

In contrast to NIOSH, which at present views the urinalysis data as its primary tool for the 

reconstruction of dose from uranium exposure, FMPC personnel considered these data as having 

a limited, if not questionable, value as given in the following statements by FMPC personnel: 

• Excerpts 

From: J.A. Quigley, MD, Director of Health and Safety 

 To: _____ of the U.S. AEC 

 Date: Nov. 1, 1963 

. . . Exposures to internal emitters is the more serious type of exposure at the 

feed materials center and probably also at the mills which prepare 

concentrates for the feed materials center . . . 

 We use urinary uranium excretion information along with air survey 

information to be sure that we are controlling airborne exposures to amounts 

that will not be harmful. We do not consider the urinary uranium excretion 

measurements as an accurate method of estimating either body burden or 

exposure. We have assumed that the determination of internal exposure by 

any method or combination of methods is less precise the [sic] are estimations 

of exposure to external radiation . . . Our urinary uranium excretion records 

substantiate this opinion. [Emphasis added.] 

• Excerpts

From: M.S. Nelson, Manager/FMPC 

To: U.S. AEC 

Date: March 6, 1972 

. . . As we have pointed out on previous occasions, we have little confidence 

in the reliability of any method for assessing dose to the lung from depleted, 
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normal, or lightly enriched uranium at the expressed level of interest, 25% to 

50% of the annual standard. We believe that uranium assay results are of no 

value for this purpose. 

 It must be assumed that anyone who works in the production area has the 

potential for acquiring 25% of a permissible lung burden. At present there 

are 353 wage employees working in the production area. Also there are 

about 50 salaried employees who have the potential for acquiring 25% of a 

permissible lung burden. [Emphasis added.] 

• Excerpts

From: S.F. Audia, Manager/FMPC 

To: U.S. DOE 

 Date: Aug. 1, 1979 

. . . Uranium urinalysis results are not used to evaluate radiation exposure 

at the FMPC. . . . Urinalysis results are only used as an indication of the 

adequacy of basic exposure control. [Emphasis added.] 

• Excerpts

From: R.M. Spenceley, Manager/FMPC 

To: Battelle-Northwest Laboratory 

Subject: Answers to a Questionnaire on Radiation Recordkeeping 

 Date: July 29, 1984 

 Among various questions asked in the 1984 Questionnaire was Request #8, which asked 

FMPC to respond to the following: 

If you do not calculate radiation dose equivalents from internally 

deposited radionuclides please describe what data you record concerning 

internal exposure and the types of analyses which may be possible to perform 

on the data . . . 

 FMPC – Response to Request #8

(1) Amount of deposited nuclide determined from lung count is recorded and 

can be used to calculate lung burned. 

(2) Excretion urinalysis data recorded but this cannot be used for 

calculating internal dose. [Emphasis added.]

When urine data for uranium exposure is either absent or incomplete, NIOSH intends to employ 

air monitoring data and/or in vivo lung count data. Findings in behalf of these two alternative 

approaches are discussed elsewhere in this report. 
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Finding 4.1-3: Failure to Monitor All Personnel with Potential Internal Exposure to 

Uranium 

In Section 7.2.1.2 of the SEC-00046 Evaluation Report, NIOSH states the following: 

Since nearly all FMPC workers were monitored for uranium in urine, no coworker analysis has been deemed necessary for uranium intakes. [Emphasis 

added.]

SC&A interprets NIOSH’s position to imply that if a FMPC worker was not monitored for 

uranium intake, there is no justification or need for assigning internal dose(s) for uranium 

intake(s) other than “environmental intakes.” 

While records suggest the episodic monitoring of many production workers, there is evidence 

that ubiquitous fugitive emissions exposed non-production workers at other locations who were 

not monitored. 

Attachment 4.1-3 identifies uranium results in behalf of four individuals even though they were 

not considered at risk and “. . . there were no apparent reasons for the high uranium results.” 

It should further be noted that FMPC had identified action levels regarding urine bioassay 

values. In a memorandum dated April 19, 1972, J.A. Quilgley, M.D., head of FMPC’s Health 

and Safety Program, the following criteria were identified: 

Urine Results. . . . Persistent results over 0.025 mg/l indicate moderate exposure 

and results over 0.040 mg/l are considered due to excessive exposures which 

require follow-up. [Emphasis added.] 

Thus, the “unexplainable” urine result of 0.543 mg/l in individual #3 is more than 13 times the 

value of 0.04 mg/l action level. Of concern to SC&A is the fact that this value was observed in 

an individual whose “. . . possibility of . . . getting an exposure of 0.543 mg U/l is very remote.” 

A reasonable interpretation of Attachment 4.1-3 is that these unexpected bioassay results may 

have involved “non-production workers” whose urine data were to have provided a baseline 

control value. If this interpretation can be substantiated, then it must be concluded that all 

workers (regardless of job-function/classification) were likely exposed internally to substantial 

amounts of uranium (as well as thorium), but may not have been monitored. 
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ATTACHMENT 4.1-3 
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Finding 4.1-4: Use of Claimant-Unfavorable Assumptions and Default Values Regarding 

the Level Uranium Enrichment

The degree of U-235 enrichment of uranium feedstocks impacts the interpretation of uranium 

bioassay measurements, which are limited to fluorophotometric values defined in units of mg of 

elemental uranium per liter urine. 

To account for the variability of uranium enrichment, Section 5.2.1.1 of the FMPC TBD 

provides the following: 

 During the following production years [after 1964], uranium was processed in a 

variety of enrichments ranging from depleted to as high as 20%. The quantities 

of enriched material above 2% was not documented but was qualitatively 

reported to be small and/or insignificant in total mass . . . 

In the absence of specific enrichment information . . . the default assumption for 

time periods after 1964 is 2% enrichment for bioassay data in milligram 

quantities of uranium. Prior to 1964 natural uranium should be used. 

[Emphasis added.] 

While SC&A concurs with the statement that “. . . quantities of enriched material above 2% was 

likely to be small/insignificant in total mass, the concluding default value of 2% enrichment is 

likely to significantly underestimate the interpretation of urine bioassay for select workers. 

Moreover, these workers cannot be separated from the rest because the bioassay was restricted to 

photofluorometry and the in-vivo counting made an across the board 1% enrichment assumption. 

Furthermore, enriched uranium was introduced into Fernald much earlier (SC&A 2006c). 

Enclosed as Attachment 4.1-4A is a 1968 Health Protection Appraisal Report for NLO, which 

includes the following statements: 

“. . . action has been initiated for handling U-235 enrichments above 5%. 

Current plans include the installation in Plant 1 of a geometrically safe 

continuous digester for enrichments to 10%. It is expected that this facility will 

be operational by March 

1969 . . . 

Projected and Anticipated U-235 Enrichment Processing

 Discussions with CAO and NLO personnel have indicated that Fernald will 

probably reprocess cold fuel from several reactor sites including Hallan, Bonus, 

EGCR, Piqua, and perhaps SRO. Significant portions of the fuel will range 

from 3% to 7% U-235 enrichment. In this regard, a campaign is scheduled to 

begin in February 1969. [Emphasis added.] 

The above-cited statements suggest that for select years and FMPC facilities, some workers may 

have been exposed to uranium materials that are significantly higher than the 2% default value. 
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ATTACHMENT 4.1-4A 

EXCERPTS FROM HEALTH PROTECTION APPRAISAL REPORT 

DATED SEPTEMBER 1968
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Attachment 4.1-4A (Continued) 
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Finding 4.1-5: There are several radionuclide contaminants in RU that are not adequately 

considered for internal dose estimates. Most relevant to this concern are impacts of these 

contaminants in RU raffinate waste streams. 

The TBD focuses on Pu-239, Np-237, and Tc-99 in its evaluation of RU. Other radionuclides, 

such as americium-241 and thorium isotopes (228, 230, and 232) are mentioned, but no data are 

provided (see Section 5.2.2, Vol. 5, pp. 13–18). The omission of thorium isotopes, and in 

particular thorium-230, may be of considerable significance. Furthermore, when RU is 

processed for its uranium content, the raffinates tend to accumulate the plutonium and other trace 

contaminants, including thorium-230. The raffinate stream contains little uranium. Hence the 

problem of dose reconstruction for workers who handled the raffinates is analogous to that of the 

workers who handled the waste streams from pitchblende ore processing. This problem was 

recognized at Fernald at least by the mid-1980s. For instance, a 1988 evaluation stated the 

following: 

The uranium feed would contain the trace of TRU impurity that was typical of 

recycle uranium. A portion of the TRU impurities would end up in the uranium 

product and a portion in the byproducts. The vast majority of uranium goes into 

the uranium product, but a small amount does end up in the byproduct. The end 

result is that the ratio of TRU to U is slightly lower in the product than it was in 

the feed, but that ratio is much higher in the byproduct than it was in the feed. 

[Hinnefeld 1988, emphasis added]. 

This problem of concentration of trace radionuclides in the raffinate stream is also recognized in 

the TBD, which cited an expert evaluation done in 1989 (Bassett et al. 1989). In the case of 

magnesium fluoride feed, a note to Table 5-9 in the TBD states the following: 

Though the results in the table are all reported in ppb U, this measure is 

meaningless in subgroups in which there is very little uranium, such as 

subgroup 8, in which the MgF2 did accumulate some isotopes, but was low in 

uranium by design. [Vol. 5, TBD, p. 15]

Despite the fact that the TBD states that trace contaminant values are “meaningless” when there 

is very little uranium present, the quantitative discussion in the TBD of RU dose estimation is 

focused primarily on the trace contaminant values of uranium feed material, rather than raffinates 

or magnesium fluoride. 

Thorium-230 has also been recognized as a specific problem in this regard. For instance, the 

DOE-commissioned evaluation of radiation doses due to trace contaminants in RU for the 

Paducah plant indicates that thorium-230 doses were among the highest in some circumstances. 

In that case, the maximum bone surface dose estimated for “ash receivers” was estimated as 

110 rem, about the same as that for Pu-239 and much higher than Np-237 (PACE/University of 

Utah 2000, Tables 7.10 and 7.11, pp. 76–77). As with the processing of ores, thorium-230 will 

tend to concentrate in the raffinate stream as well, exacerbating the problem. 
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A complete evaluation should also consider thorium-232, uranium-232, and uranium-236 as 

potential contaminants of RU. Specifically, U-232 is created as a neutron activation (decay) 

product of protactinium-231, and the DOE recommends that it be taken into consideration in RU 

assessments. This is both an internal and external dose issue, because U-232 decays into 

thorium-228 with a 70-year half-life. According to DOE-STD-1136-2004, the Guide of Good 

Practices for Occupational Radiological Protection in Uranium Facilities, U-232 presents the 

greatest external dose hazard in RU: 

The isotope in recycled uranium presenting the greatest potential radiological 

hazard from external sources is 232U. 232U is a daughter product of neutron 

activation of 231Pa. The health hazards of 232U are primarily due to the rapid 

buildup of gamma activity of its decay products, particularly from 228Th. The 

gamma activity buildup is both time and process-dependent. [DOE 2004, 

p. 2-15]. 

Given that the highest values of trace contamination with plutonium received at Fernald could be 

in the thousands of ppb, Table 5-9, which gives the values for plutonium in the various RU 

streams at Fernald, is incomplete and inadequate. The highest value of Pu-239 in this table, 

412.177 ppb, corresponds to feed material (ash) from Paducah. There is no discussion in the 

TBD of the specific batches of RU and the waste streams arising from them. Such an analysis is 

necessary for individual internal dose reconstruction for at least some groups of Fernald workers. 

Finding 4.1-6: The data on trace contaminants in RU in the Fernald TBD are incomplete 

and appear to be incorrect. Different official documents have very different values for 

various aspects of RU data, including production and contamination. The contradictions 

have not been sorted out in the TBD. (Note: This finding was previously identified in 

SC&A 2006c.) 

The TBD cites considerable data on the contamination of RU with trace amounts of 

plutonium-239, neptunium-237, and technetium-99. However, these data are incomplete. The 

representation of maximum trace contamination is at variance with other official documents and 

appears to be incorrect. 

Table 5-9 of the TBD (Vol. 5, pp.15–16) provides data on plutonium-239, neptunium-237, and 

technetium-99 contamination of various sources of RU received at Fernald. The data are given 

in parts per billion of the trace contaminant in uranium, written as “ppb U,” which we will 

abbreviate here simply as ppb. The highest value of Pu-239 contamination, associated with 

uranium trioxide from tower ash from Paducah is given as 412.177 ppb (TBD, Vol. 5, Table 5-9 

and p. 17). Several values of Pu-239 contamination are between 10 and 100 ppb, and the rest are 

below 10 ppb, which was the specification limit for Pu-239 contamination at Fernald. 

However, other documents are at variance with the maximum value of 412.177 ppb. For 

instance, a 1985 compilation of RU feed materials above 3 ppb U received at Fernald, prepared 

by National Lead of Ohio (NLO), gives the highest total plutonium contamination in “ash” for 

1980 as 1,122.553 ppb (Spenceley 1985). Plutonium isotopes other than Pu-239 would not 

contribute significantly to the ppb values; hence, it is safe to interpret the “Total plutonium ppb” 
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in Spenceley 1985 as approximately equivalent to a Pu-239 ppb value. This reference does not 

provide details of other radionuclides. 

It is quite possible that both sources are incorrect. The TBD appears to be based on a DOE 

report on RU (DOE 2000). This DOE report states that the total uranium receipts at Fernald 

amounted to 362,581.8 metric tons (DOE 2000, p. ES-2). This appears to be inconsistent with 

materials accounting reports from Fernald. For instance, the cumulative receipts until the end of 

FY 1986 were stated by Westinghouse to be 606,931.9 metric tons (Bogar 1986, Table V). This 

materials account is consistent with others produced during the period of production and 

submitted to the AEC and the DOE. Hence, it is likely that the DOE 2000, which is the basis for 

the data on RU, is incorrect even for the basic value relating to uranium receipts at Fernald. 

We note here that the TBD also appears to have an incorrect value for uranium production at 

Fernald. Volume 1 of the TBD estimates the shipments of uranium metal at 170,000 metric tons 

and intermediate products at 35,000 metric tons. The materials account cited above provides a 

value of 594,699 metric tons cumulative shipments to the end of FY 1986. Furthermore, the 

total shipments of 205,000 metric tons in Volume 1 of the TBD are less than the estimate of 

246,683 metric tons of RU alone that the TBD states were received at Fernald (TBD Vol. 5, 

p. 13). The total amount of RU of 246,683 metric tons in DOE 2000 appears rather large. In 

contrast, the amount estimated in Spenceley 1985 is only 7,183.6 metric tons, cumulative 

through 1985. This is almost 30 times less than the value in DOE 2000. Finally, Volume 6 of 

the TBD contains an entirely different number for RU compared to Vol. 5. Citing a DOE 2003 

report on RU, it states that the receipts of RU at Fernald amounted to 17,966 metric tons. This 

matter is further discussed in Chapter 7. 

Since the last mentioned report (DOE 2003) was prepared in order to correct “some 

inconsistencies between quantities of RU shipped and the quantities received” (DOE 2003, p. v) 

in the DOE 2000 report, it is surprising that NIOSH did not employ the corrected report. The 

figures from this report are shown in Vol. 6 of the TBD (Table 6-2, p. 8); they indicate an overall 

average Pu contamination of the RU received at Fernald of 4.14 ppb. This is greater than all the 

average contamination values for enriched, natural, and depleted RU shown in Table 5-10, Vol. 5 

of the TBD (3.5 ppb, <0.1 ppb and <0.1 ppb, respectively), where the basis for the RU dose 

reconstruction is developed. Volume 5 of the TBD gives the overall average Pu contamination 

of RU as 0.9 ppb, which is only about 22% of the value in DOE 2003. 

The overall average values for Np-237 and Tc-99 in Table 5-10 are also at variance with DOE 

2003 (Table A-9, p. 60, which is reproduced as Table 6-2 in Vol. 6 of the TBD). The 

concentrations of Np-237 and Tc-99 calculated from the DOE 2003 data are 319 ppb and 

7510 ppb, respectively, compared to 104 ppb and 1,346 ppb given in Vol. 5 of the TBD. 

Another contradiction emerges from the comparison of the values in Volume 5 of the TBD and 

DOE 2003. Since the overall value for RU receipts given in Volume 5 of the TBD is so much 

larger than that in DOE 2003 (246,683.1 metric tons versus 17,966 metric tons), the total 

contaminant content of RU estimated in Volume 5 of the TBD is much larger than that in DOE 

2003. Table 4.1-1 shows the comparison. 
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Table 4.1-1. Comparison of Total Fernald RU Contamination, TBD, Vol. 5 vs. DOE 2003 

Total Pu, grams Total Np-237, grams Total Tc-99, grams 

TBD Vol. 5, Table 5-10 217.7 25,742.1 331,998.1 

DOE 2003, Table A-9 74.3 5,735 135,000 

Ratio, TBD/DOE 2003 2.93 4.49 2.46 

 

Volume 5 of the TBD is not only in contradiction to Vol. 6 of the TBD, but it is also at variance 

with what DOE claims is a more definitive DOE report on RU (DOE 2003), since that report was 

designed to correct earlier problems. SC&A has not done a review of the underlying data that 

led to the correction, since that would involve a major effort to review RU data across the 

complex. It is unclear at the present time how reliable the various figures for plutonium, 

neptunium, and technetium contamination (both as totals and in terms of concentrations) might 

be. 

An evaluation by Bechtel of RU shipped to Fernald from Paducah provides yet another set of 

values for RU contamination for a specific batch that do not match the values in Volume 5 of the 

TBD. This document provides a range of values of Pu-239 in “Feed Plant Ash” shipped to 

Fernald in 1980 from Paducah as 37 to 3,118 ppb (Bechtel 2000, Table 4-2.2, p. 51). The 

various containers of Feed Plant Ash were not mixed at Paducah, because the reported values 

were “calculated from results of 16 hoppers analyzed by FMPC” (Bechtel 2000, footnote to 

Table 4.2-2, p. 51). Hence workers handling and measuring the hoppers would have been 

exposed to concentrations of trace radionuclides during some time periods when the 

concentrations were far higher than the largest value reported in the TBD, especially if they 

worked with raffinate streams. Finally, the same table in Bechtel 2000 notes that data for 

plutonium contamination are not available for “Filter Cake” shipped to Fernald. Three other 

items have only blanks in the column for plutonium contamination. The higher trace 

contamination levels may adversely affect some workers, who were mainly in contact with them 

due to their work assignments or periods of employment. For others, long-term exposure to RU 

may mean that the values suggested by NIOSH are claimant favorable (see below). 

In summary, the contradictions and conflicting values in the RU data need to be investigated, 

before a reliable set of values for RU amounts and contamination can be established. 

4.2 FINDINGS ASSOCIATED WITH DOSE ASSESSMENTS IN BEHALF OF K-65 

WASTES AND OTHER RAFFINATE WASTES 

Finding 4.2-1: The K-65 Default Model is Inappropriate

Section 7.4.1.3 of the ER addresses the petitioner’s concern about worker dose assessment in 

behalf of K-65 processes involving Silos 1 and 2. For K-65 processes, NIOSH states that a 

maximized approach will be employed as described on page 27 of ORAUT-TKBS-0017-5. 

SC&A has evaluated the proposed model and concludes that the model contains numerous 

assumptions that are inappropriate and consistently non-claimant favorable. 
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Key Elements of the K-65 Dose Model

Critical parameters that define NIOSH’s K-65 Silo dose model include the following empirical 

data: 

• Internal and external exposures were received by workers who transferred 13,000 drums 

of K-65 wastes into Silos 1 and 2 between July 1952 and September 1958. 

• The isotopic composition of K-65 wastes in Silos 1 and 2 are given in Table 5-16 of the 

TBD and reflect a 1993 study of core samples. 

• A single record dated 4-15-1953 identifies the transfer of an “. . . average about 80 

drums/day.” 

• A small number of records/data sheets from 1952 and 1953 involve air samples with a 

wide range of activity levels defined in alpha activity per cubic meter of air. Values 

ranged from <1 MAC to 18,777 dpm/m of 268 MAC. 

• Air sampling consisted of both general air (GA) and breathing zone (BZ) samples. The 

flow rate of air samples was consistently recorded at 0.02 m3

/min (or 20 liters/min) for 

both GA and BZ sampling with sampling times ranging from 1 to 30 minutes. 

While these limited empirical data are helpful in the development of a bounding internal dose 

model, these data provide no concrete information regarding the amount of time a given worker 

was exposed to these air contaminants in any given year. To fill these gaps of information, 

NIOSH’s dose model is based on the following assumptions: 

(1) The transfer of 13,000 drums was an around-the-clock effort that involved three shifts 

of workers at 80 drums per day. 

(2) A group of external dose data sheets (i.e., film badge records) for 22 K-65 workers were 

used as the basis for defining the yearly exposure duration to K-65 airborne 

contaminants and include the following: 

• Of the 22 K-65 workers (identified in a small number of data sheets that spanned 

only a brief time period), NIOSH chose 13 workers with the highest doses ranging 

from 158 to 500 mrem average per week. Available records show that three of 

the 13 workers were assigned to K-65 for three weeks; and there were 10 workers 

with K-65 dosimetry data for 6 weeks. The highest recorded weekly external 

gamma dose among the 13 workers was 1,200 mrem. 

• The collective average exposure for all 13 workers was calculated at 

312 mrem/week. 
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This collective average external dose of 312 mrem/week for the 13 highest K-65 workers was 

used by the model to justify the yearly exposure time to K-65 airborne levels by means of the 

following assumptions: 

• NIOSH assumed that for 1952 the annual external exposure limit for penetrating 

radiation was 5 rem. 

• NIOSH further assumed that FMPC must have had a more restrictive administrative 

dose limit of 4 rem/year. 

• By dividing the assumed administrative dose limit of 4 rem/year by 312 mrem/week, 

NIOSH concluded that K-65 workers would be restricted to a maximum of 3 months

after which the workers would have been shifted to a non-radiological work location. 

• The above-derived 3 months per year exposure duration was further reduced to 6 weeks, 

as explained by the following statement on page 27 of the TBD: 

. . . From the information derived in the external dose data sheets and the air 

monitoring sample sheets, it appears that the transfer could have been limited 

to a period of 10 weeks per year with no individual working more than a 

period of 6 weeks in the year in order to control external dose within the 

regulatory limits.

Thus, NIOSH’s final reduction of the yearly internal exposure time to 6 weeks was based on 

the assumed administrative dose limit of 4 rem/yr and the highest recorded weekly dose of 

1,200 mrem among the 13 highest externally exposed K-65 workers. 

SC&A concludes that NIOSH’s internal dose model for K-65 workers is based on incorrect and 

unfounded assumptions that are consistently unfavorable, as explained below. 

• Use of the 13 Highest Externally Exposed Workers. By using the highest externally 

exposed workers (and applying their exposure data to regulatory/administrative dose 

limits), NIOSH has employed a modeling approach that is not only irrelevant to internal

exposure but is non-favorable to the claimant. 

• Exposure Limits. During this 6-year period, the exposure limit employed by the AEC

was 0.3 rem/week, 3.9 rem/13 weeks, and 15 rem/year, which is three times higher than 

NIOSH’s “assumed” value of 5 rem/year. It was not until 1958 that the 5 rem/year limit 

was introduced in combination with the lifetime limit defined by 5(n-18) (see 

Attachment 4.2-1A). 

• Administrative Dose Limit. Review of available data provides no indication that FMPC 

imposed an administrative dose limit of 4 rem/yr. In fact, three separate data sheets 

identified as Attachments 4.2-1B, 4.2-1C, and 4.2-1D reference the AEC dose limit of 

300 mrem/week. Attachment 4.2-1D identifies the maximum permissible monthly dose 

of 1,250 mrem (or 15 rem/yr) for whole-body gamma as late as 1959. Moreover, data 
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sheets/memos show that dose limits were routinely exceeded as well as ignored

(Attachment 4.2-1D). 

• Three-Shift Work Schedule. Support for the assumption of a three-shift work schedule 

for all 6 years of K-65 operation is based on a single data sheet dated October 16, 1952. 

This data sheet identifies the shift schedule and personnel at the “K-65 Dumping 

Station.” 

It must be recalled that in addition to the 13,000 drums received from MCW, Silos 1 and 

2 received raffinate wastes produced at FMPC, which were “transferred via pipes to 

Silos 1 and 2. It is unclear whether individuals cited in the October 16, 1952 data sheet 

were assigned to drum disposal activities or to assuring that FMPC-produced raffinates 

were properly disposed. 

The likelihood that these workers were assigned to FMPC-produced raffinate disposal on 

a three-shift basis is based on the fact that processing of uranium ores at FMPC was a 

24-hour operation. 

It is also counter-intuitive to assume that the manual removal and disposal of MCWproduced raffinate waste would be performed during hours of darkness. Moreover, there 

is no logical basis to suggest an urgency to compress disposal into a 6-week period on a 

three-shift schedule, and then abandon this task for the balance of the year. 

In concluding remarks, NIOSH describes that mode as one that is “. . . based upon assumptions 

that are cumulative conservative, claimant favorable, and establishes an upper bound of 

intake for workers involved in the transfer operation of the 13,000 barrels of the stored 

MCW raffinates to the K-65 silos.” [Emphasis added.] 

In contrast, SC&A regards NIOSH’s dose model as one that is based on incorrect assumptions, 

unsupported assumptions, and counter-intuitive assumptions that are consistently claimant 

unfavorable. 
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ATTACHMENT 4.2-1A 
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ATTACHMENT 4.2-1B 
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ATTACHMENT 4.2-1C 
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ATTACHMENT 4.2-1D 
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Attachment 4.2-1D (Continued) 
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Attachment 4.2-1D (Continued) 
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Finding 4.2-2: Neither monitoring data nor a default dose model exist for estimating 

internal dose from raffinate streams associated with high-grade ore processing in Plant 2/3. 

(Note: This finding was previously identified in SC&A 2006c.) 

Fernald processed high-grade ores, including pitchblende, which was processed during the 1953 

to 1955 period (TBD Vol. 5, p. 7). These ores, being very rich in uranium (up to two-thirds 

uranium oxide content), therefore also have high concentrations of Ra-226 and Th-230, which 

are decay products of U-238 generally present at levels close to equilibrium with U-238. 

Similarly, they have relatively high concentrations of protactinium-231 and actinium-227 (and its 

decay products, thorium-227 and radium-223), all of which are in the decay chain of 

uranium-235. Processing of high-grade ores gives rise to waste streams that are high in the 

decay products of U-238 and U-235, but relatively low in uranium, which is part of the product 

stream. 

Essentially no personnel monitoring for the decay products of U-238 and U-235 was done in the 

period of production when ores were handled at Fernald. The TBD cites some air concentration 

data for Plant 2/3 (Vol. 2, pp. 21–22), but these are in production areas, not waste stream areas. 

Unlike production areas, where uranium bioassay data can provide at least a starting point for 

internal dose reconstruction, such data are not very useful in determining dose of decay products. 

Uranium is a minor constituent of the waste streams in terms of its fraction of the total 

radioactivity per gram of material. This is accentuated by the fact that the DCFs for most organs 

of the trace constituents are much larger than they are for any of the isotopes of uranium present 

in natural uranium. 

The TBD has data on the isotopic composition of the waste streams as present in an aggregated 

form in the K-65 Silos, also called Silo 1 and Silo 2. These data can be used to estimate doses in 

the absence of personnel monitoring data, provided sufficient air concentration data are 

available. However, no such data are cited for the processing waste streams in Plant 2/3. The 

TBD cites the decay products that are the radionuclides of concern for Building 3E, where 

raffinates were processed. However, there are no data that would be useful for estimating doses 

due to these radionuclides in the TBD. 

The problem of estimation of doses for production workers who worked at the filter presses and 

other locations where the waste streams were handled was dealt with by SC&A at length during 

consideration and review of the MCW SEC Petition (1949–1957 period). NIOSH also 

considered it in detail as part of that same process. The analyses and reviews can be found in 

SC&A 2005a and SC&A 2005b. The dose reconstruction procedures suggested by NIOSH, as 

well as illustrative examples, are in the attachments to SC&A 2005b. A review of those 

procedures can also be found in SC&A 2005b. 

While some of the analysis in the MCW-related reports is specific to that site, given that residues 

sent to storage were brought back to the site and reprocessed for uranium extraction, the primary 

discussion relating to pitchblende waste streams applies here and will not be repeated. Suffice it 

to say that uranium bioassay data provide an uncertain basis for estimating internal dose, due to 

the trace constituents in pitchblende processing waste streams. For such data to be used at all, 

knowledge is needed of the fraction of uranium relative to the other radionuclides at various 

points in the waste stream. Furthermore, these data need to be rather reliable, because the dose 
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depends greatly on accurate (or upper-bound) knowledge of the ratio of trace radionuclide 

activity to uranium activity. No relevant data specific to Fernald are provided in the TBD, and 

SC&A has not come across any in the course of this review. 

One alternative approach that NIOSH suggested in the context of Mallinckrodt was to use radon 

breath data. Such a dataset was available for a subset of Mallinckrodt workers. The Fernald 

TBD mentions that “a series of radon breath samples” were located in the context of a discussion 

of the composition of the K-65 silos (TBD Vol. 5, p. 26); however, the data are not provided, nor 

is it clear whether any of the workers who processed the waste streams in Plant 2/3 were covered 

by the sampling program. 

As it stands, the TBD has no procedure in place and no data on which to base doses to workers 

involved with ore processing waste streams. It is to be noted in this context that one of the 

statements made by SC&A in the Mallinckrodt context was that using the general approach for 

estimating maximum plausible doses using ORAUT-OTIB-0002 may result in doses smaller 

than the ones actually experienced by some workers. 

Finding 4.2-3: Incorrect Model Assumptions Pertaining to Radon Releases from K-65 Silos 

MCW and FMPC raffinates disposed and stored in Silos 1 and 2 at FMPC contained large 

amounts of Ra-226 that served as source term for the Rn-222 emissions that would have exposed 

K-65 workers, all other FMPC workers, as well as members of the general public. 

Radon exposures from Silos 1 and 2 are based on modeled data as described in a 1995 study 

(RAC 1995). Key information used in this study were the isotopic composition of Silos 1 and 2, 

as summarized in Table 5-16 of the TBD and restricted release rates through structural fissures 

that are the result of diurnal barometric pressure changes. 

In Section 5.2.4 of the TBD, NIOSH states that: 

. . . during the 1953 to 1978 period 5,000 to 6,000 Ci/year of 222Rn were released 

from the silos (RAC 1995). Considering the expected large differences in release 

rates due to barometric pressure changes, the release rates would average up to 

15 to 20 Ci/day after addition to the Silos were complete. 

NIOSH’s ER of SEC-00046 further states the following: 

. . . In addition to Section 5.2.4 of Technical Basis Document for the Fernald 

Environmental Management Project (FEMP) - Occupational Internal Dose, The 

Pinney study, which estimated exposure to all FMPC workers, may also be used 

to bound the doses to workers when sampling results are not available (Pinney, 

2004).

SC&A has reviewed key assumptions of the RAC 1995 model and concludes that these release 

rates are a factor of 10 to 20 too low. Our conclusion is based on the fact that the RAC exposure 

model is not based on empirical data, but represents a conceptual construct of restricted release 
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rates driven by diurnal variations in atmospheric pressure. While the model may be credible for 

a restricted time period that approximates the time periods of the 1995 study, the basic model 

assumption of a restricted release driven by diurnal barometric pressure is inappropriate for 

periods that predate 1986, as explained below. 

A letter dated September 13, 2006, to one of the SEC-00046 petitioners provided the following 

information: 

The K-65 Silos . . . were constructed in 1951 to 1952, and materials were added 

to the Silos from July, 1952 to September, 1958. The Silos had problems with 

deterioration almost since the time of construction. Significant cracking in the 

walls and seepage were noted from the 1950s. Because of these problems, 

periodic repairs and improvements to the Silos were implemented from the 1960s 

through the 1980s. 

 . . . Placement of protective dome covers in 1970s, constructed of steel 

and plywood, blanketed the center of each silo. By themselves, the covers offered 

little if any barrier to the diffusion and ventilation of radon. Vents in these domes 

were sealed in 1979. In 1986, a waterproof protective membrane of liquid 

neoprene was added, and in 1987 a layer of polyurethane foam was placed on 

dome surface in an attempt to mitigate the migration of radon gas to the 

environmental. No routine environmental monitoring for radon was performed 

prior to 1980. [Emphasis added.] 

On the basis of these modifications to the silos, the model assumptions employed in the RAC 

1995 study cannot be assumed applicable before 1986. 

Prior to 1986, a more appropriate and claimant-favorable approach would assume an unfiltered

release of radon that is unaccounted for in the isotopic composition, as defined in Table 5-16 of 

the TBD. For Silo 1, Table 5-16 identifies the following activity levels from which radon 

releases can be estimated by means of first principles: 

 Ra-226 – 477 nCi/g 

 Po-210 – 281 nCi/g 

 Pb-210 – 202 nCi/g 

First principles dictate that if no radon escaped from the raffinate wastes, the activity 

concentration of Ra-226 would be in equilibrium and, therefore, equal to those of Po-210 and 

Pb-210. Conversely, the observed disequilibrium implies the release of radon. On the 

assumption that Silo 1 contained one-half of the 10,000 metric tons of raffinate, SC&A 

calculates a radon release of (1) 64,500 Ci/yr based on the disequilibrium between Ra-226 and 

Po-210; and (2) 92,000 Ci/yr based on the disequilibrium between Ra-226 and Pb-210. 

These derived values are about 10 to 18 times higher than those estimated in RAC (1995). 

SC&A favors the higher value of 92,000 Ci/yr since Pb-210 has a 21-year half-life and is less 

affected by potential ingrowth than Po-210. (The improved dome cap modifications in the 1980s 
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may have increased retention of Rn-222, which in turn may have differentially increased the 

ingrowth of Po-210 above that of Pb-210.) 

4.3 FINDINGS ASSOCIATED WITH INTERNAL DOSE ESTIMATES FOR 

THORIUM 

NIOSH’s Changing Position on Proposed Approach for Reconstructing Thorium Doses

At this time, there is considerable uncertainty with regard to the approach NIOSH intends to 

employ for estimating internal exposures to thorium for “maximized” and “best estimate” dose 

reconstructions. 

Proposed Method in TBD Volume 5. Upon review of available monitoring data for thorium, 

NIOSH initially concluded the following, as stated in Section 5.3.1 of the FMPC TBD (ORAUTTKBS-0017-5): 

 From pages 22 and 23: 

A fundamental difficulty of dose reconstruction for thorium processing is that 

either 1) in vitro bioassays for thorium were not performed or 2) data is not 

available until after 1986. An additional consideration is that air sampling data 

was not used to calculate intake and dose until after 1986. Air monitoring was 

used only to control exposures to levels below the MAC. A number of internal 

memoranda identify those areas with concentrations at or above the MAC of 

100 dpm m3 (4.5 ×10-11 μCi cm3) as areas requiring respiratory protection. 

However, recorded examples of exposure to multiple MAC levels without 

respirators indicate these violations of policy were not uncommon. In addition, 

the urine sampling was performed for uranium only. The only discovered record 

of thorium exposure has been in vivo lung count data sheets in a few claimant 

records and a single claimant record which indicates thorium urine results, 

counted for beta and at essentially no detectable results, from before 1986.

Thorium processing was completed in 1979, with exposure from that time being 

limited to repackaging and shipping operations. 

After 1986 thorium air sampling was used to estimate internal exposure using 

continuous lapel air samples as breathing zone (BZ) evaluations. From that time 

until the present air monitoring is used to conservatively estimate internal intake 

even when the worker wore respiratory protection.

Based upon the above information and assumptions, the recommended claimantfavorable default exposure approach to assign thorium intakes is to assume: 

• An intake for an exposure period of 100 hours per year at an assumed 

exposure of 10 MAC is judged adequate to account for the higher levels of 

exposure indicated by air sampling, since few samples above 10 MAC were 

reported and these primarily represented short term maximized sampling 
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(based upon descriptions on the sample sheets). Also typical and more 

extensive uranium air sample data demonstrate that 10 MAC is a reasonable 

assumption of the higher level of exposure. 

• No respiratory protection factor, although not wearing respirators when air 

concentrations were above MAC represented procedural violations. This 

violation was known to have occurred and was not unusual. 

• An intake for an exposure period of 500 hours per year at an average air 

activity of 0.1 MAC during normal operations 

Using these assumptions, the claimant-favorable assumption would be:

500 hr × 0.1 MAC + 100 hr x 10 MAC = 1050 MAC-hr exposure (5-1) 

. . . This default intake rate applies to the most exposed craft personnel (e.g., 

chemical operators, process maintenance personnel, safety personnel, and first 

line supervisors/ foremen) at the locations and during the periods noted below. 

For workers whose location cannot be determined the claimant favorable 

assumption is that they were exposed to thorium and daughter products when they 

were employed during the listed periods in Table 5-16. Exposures to casual 

workers who worked in the immediate vicinity of the plants should be evaluated in 

the Environmental Occupational Dose section. 

Thorium chest counts may indicate that lower exposures occurred. When 

available, the chest count data should be used to constrain an employee’s intake. 

In other words if there are chest count results for an employee, the smaller of the 

default thorium intake or the chest count determined intake should be assigned for 

full dose reconstructions. [Emphasis added.] 

Proposed Method in SEC-00046 Evaluation Report. In the SEC-00046 ER, issued October 25, 

2006, NIOSH modified its position, as given in Section 7.2.1, by the following statements: 

NIOSH reviewed available process information for the entire operational period 

of the proposed class. Air monitoring programs were in place during this entire 

time period. These programs covered all operational areas and emphasized 

sample collection in process areas with higher potential for airborne 

contamination. In addition to general area airborne concentration levels, there 

are data available from job-specific breathing zone air sampling events.

. . . Air monitoring specifically identified for thorium operations, coupled with an 

extensive lung count database during 11 years of that period, provides a basis 

for default intakes for workers who worked with these materials during the 

recorded periods of operation. [Emphasis added.] 
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In Section 7.2.3 of the ER, entitled Internal Dose Reconstruction Feasibility Conclusion, NIOSH 

states the following: 

Recorded bioassay for uranium is extensive and sufficient for internal dose 

reconstruction. Thorium bioassay is not as well documented. However, with the 

addition of a database of in vivo bioassay (lung counting) and additional air 

sampling data currently being expanded, the air monitoring database is being 

strengthened through recent retrieval of additional field data, which will be used 

to ensure a claimant favorable dose reconstruction to represent maximum dose.

[Emphasis added.] 

At a meeting of the Advisory Board held in Mason, Ohio, on February 8, 2007, NIOSH 

presented its evaluation of the SEC Petition-00046 and further refined its approach for assessing 

thorium exposures. Presented as Slide #10 and shown here as Exhibit 4.3-1 is NIOSH’s most 

current proposed method for assessing thorium doses. 

EXHIBIT 4.3-1 

SC&A interprets Exhibit 4.3-1 to imply that NIOSH intends to develop a model that correlates 

breathing zone air sample data (taken at various FMPC locations over time) with worker-specific 

lung counts. SC&A also assumes that this thorium dose model will be used as follows for 

assigning intakes: 

(1) A claimant will be identified by job function/FMPC work location. 

(2) Based on job function/FMPC work location, NIOSH will assign a breathing zone air 

concentration for the duration(s) of thorium exposure work period(s). 
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(3) Based on NIOSH’s correlation study of “6000 in vivo results for thorium (Ac-228 and 

Pb-212), with 2000 – 4000 thorium air sample results,” NIOSH intends to assign a 

claimant-specific thorium intake that is linked to the claimants’ job title, assigned work 

location, and employment period. 

The credibility and feasibility for assigning thorium intakes by means of these approaches are the 

subject of Findings 4.3-1 through 4.3-10, as discussed below. 

Finding 4.3-1: Generic Limitations and Uncertainties Associated with Air Sampling for 

Monitoring Worker Exposures 

There are two basic kinds of air sampling strategies used to monitor worker exposure: fixed 

station, commonly called general area (GA) sampling, and breathing zone (BZ) sampling. 

Because sources of airborne contamination are highly localized, concurrent measurements by 

GA and BZ frequently differ by orders of magnitude. 

In a 1967 study conducted at the Nuclear Materials and Equipment Corporation (NUMEC), BZ 

lapel air sampling measurements were compared to GA samples. Figure 4.3-1 illustrates the 

strong tendency of GA samples to underestimate the air concentrations to which a worker is 

exposed. Important to note is that this discrepancy increases with air concentrations and at the 

MPC level is on average about 70-fold too low (see Figure 4.3-2). 
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Figure 4.3-1. Comparison of Lapel to Fixed Station Air Sampling 

(Source: Caldwell et al. 1967) 

In spite of the intent by NIOSH to limit its use to BZ air samples, the high variability of BZ 

samples with time severely limit their use. To illustrate this variability for a fixed location and 

a fixed job task, BZ sampling data are presented in Attachments 4.3-1A to 4.3-1E that 

correspond to time differences of only minutes, weeks, and years between BZ air samples: 

• Attachment 4.3-1A: Variability Over Short Time Intervals Taken on November 19, 1970

Identified in Attachment 4.3-1A are three job descriptions. In turn, for each job task, 

three separate BZ samples were taken. For example, the first task involved an operator 

removing ThF4 from drying pans with a metal scoop, etc. Of only three BZ air samples, 

which were assumedly taken over a relatively short time, air concentrations varied nearly 

26-fold from a high of 4400 dpm/m3 to a low of 170 dpm/m3. 

For the second task, the operator who “takes full retort pans of ThF4 from ventilated 

enclosures . . ., the air concentrations varied more than 33-fold (i.e., from 100,600 to 

3,000 dpm/m3

); and for the third task that involved the “. . . operator sawing derbies, 

loading and unloading was . . .” the air concentration for three samples varied from 

45,570 to 620 dpm/m3 or more than 72-fold. 

• Attachment 4.3-1B: Defines Air Concentrations for Eight Discrete Locations at 

30-Minute Time Intervals 

For each of the eight locations, air sampling measurements were taken at 30-minute time 

intervals. For example, at the “West Separation Booth Area,” the air concentrations 

were as follows: 

 

Time Air Concentration (dpm/m3)

9:05 a.m. 355 

9:35 a.m. 140,012 

9:50 a.m. 3463 

• Attachment 4.3-1C: Variability in Air Concentrations Over a One-Week Period

 Attachment 4.3-1C cites variation in air concentrations over a one-week period. For 

example, for the 21 BZ air samples taken for workers “repairing inside furnace,” air 

concentrations varied from 81,470 dpm/m3

 to 43 dpm/m3 or nearly 1900-fold. 

• Attachment 4.3-1D: Compares BZ Data for Select Operations/Locations

 For the years 1960, 1961, and 1962, BZ air concentrations were compared for select 

operations and locations. For example, the task of “unplugging furnace discharge line” 

yielded an exposure of 417 MAC (or 29,190 dmp/m3

) in 1962, but the same operation in 

1961 resulted in an exposure of 4 MAC (or 280 dmp/m3). 
Effective Date: 

June 27, 2007 

Revision No. 

1 – Draft 

Document No. 

SCA-SEC-TASK5-0056 

Page No. 

Page 54 of 142 

NOTICE: This document has been reviewed for Privacy Act information and has been edited accordingly. 

NOTICE: This report is pre-decisional and has not been reviewed by the Advisory Board for factual accuracy or 

applicability within the requirements of 42 CFR 82. 

While spot air samples (with sampling times of only minutes) can be expected to show a higher 

level of variability, a more consistent measure of air concentration is expected when air 

concentrations are defined as daily weighted air concentrations. However, even when 

concentrations were normalized and expressed as daily weighted air concentrations, this 

variability persisted as illustrated in Attachment 4.3-1E. Attachment 4.3-1E compares the daily 

weighted thorium air concentrations for two time periods in 1955: May 17–October 31 versus 

November 4–November 23). Job descriptions showing the highest variability included the 

following: 

Daily Weighted Air Concentration (x MAC) Job Description May 7–Oct. 31 Nov. 4–Nov. 23

Wet Area Operators 215.1 2.7 

Reduction Charge Helpers 233.7 3.5 

Secondary Welder Helpers 685.6 122.1 

Primary Arc Furnace Operators 473.0 23.3 

In addition to data shown in Attachments 4.3-1A to 4.3-1E, SC&A reviewed a large body of BZ 

air sampling data, which showed variations that are wholly consistent with those of the enclosed 

Attachments. SC&A concludes that this high degree of variability of spot air samples is but one 

of several major deficiencies that severely limit their use for dose reconstruction.