Draft

ADVISORY BOARD ON

RADIATION AND WORKER HEALTH

National Institute for Occupational Safety and Health

A REVIEW OF NIOSH’S PROGRAM EVALUATION REPORT 

DCAS-PER-061: BRIDGEPORT BRASS COMPANY

Contract No. 211-2014-58081 

SCA-TR-2017-PR005, Revision 0 

Prepared by

John Mauro, PhD, CHP 

Stephen L. Ostrow, PhD 

Nicole Briggs

Robert Anigstein, PhD 

SC&A, Inc.

2200 Wilson Boulevard, Suite 300 

Arlington, Virginia, 22201 

Saliant, Inc.

5579 Catholic Church Road 

Jefferson, Maryland 21755 

July 2017 

DISCLAIMER

This is a working document provided by the Centers for Disease Control and Prevention (CDC) technical 

support contractor, SC&A for use in discussions with the National Institute for Occupational Safety and 

Health (NIOSH) and the Advisory Board on Radiation and Worker Health (ABRWH), including its 

Working Groups or Subcommittees. Documents produced by SC&A, such as memorandum, white paper, 

draft or working documents are not final NIOSH or ABRWH products or positions, unless specifically 

marked as such. This document prepared by SC&A represents its preliminary evaluation on technical 

issues.

NOTICE: This report has been reviewed to identify and redact any information that is protected by the

Privacy Act 5 U.S.C. § 552a and has been cleared for distribution.
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SC&A, INC.: Technical Support for the Advisory Board on Radiation and Worker 

Health Review of NIOSH Dose Reconstruction Program

DOCUMENT TITLE: A Review of NIOSH’s Program Evaluation Report DCAS-PER-061: 

Bridgeport Brass Company

DOCUMENT NUMBER/

DESCRIPTION:

SCA-TR-2017-PR005 

REVISION NO.: 0 (Draft)

SUPERSEDES: N/A

EFFECTIVE DATE: July 10, 2017

TASK MANAGER: John Mauro, PhD, CHP [signature on file]

PROJECT MANAGER: John Stiver, MS, CHP [signature on file]

DOCUMENT 

REVIEWER(S):

John Stiver, MS, CHP [signature on file]

Kathy Behling [signature on file]

Record of Revisions

Revision 

Number

Effective 

Date Description of Revision

0 (Draft) 7/10/2017 Initial issue
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TABLE OF CONTENTS

Abbreviations and Acronyms ......................................................................................................... 4

1.0 Statement of Purpose .......................................................................................................... 6

2.0 Relevant Background Information ...................................................................................... 8

3.0 Subtask 1: Identify the Circumstances that Necessitated the Need for DCAS-PER-061 . 10

4.0 Subtask 2: Assess NIOSH’s Specific Methods for Corrective Action ............................. 11

4.1 Internal Dose Reconstruction During Operations ................................................. 11

4.1.1 Revision 00 to ORAUT-TKBS-0030 ........................................................ 11

4.1.2 Revision 02 to ORAUT-TKBS-0030 ........................................................ 13

4.2 External Dose Reconstruction During Operations ................................................ 15

4.2.1 Revision 00 to ORAUT-TKBS-0030 ........................................................ 15

4.2.2 Revision 02 to ORAUT-TKBS-0030 ........................................................ 17

4.3 Remediation/Residual Period................................................................................ 21

4.4 Occupational Medical Dose .................................................................................. 23

5.0 Subtask 3: Evaluate the PER’s Stated Approach for Identifying the Universe of 

Potentially Affected Dose Reconstructions.................................................................................. 24

6.0 Subtask 4: Conduct Audits of a Sample of Dose Reconstructions Affected by 

DCAS-PER-061 ............................................................................................................................ 24

7.0 References ......................................................................................................................... 25
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ABBREVIATIONS AND ACRONYMS

Advisory Board

or Board Advisory Board on Radiation and Worker Health 

AEC Atomic Energy Commission

AP anterior-posterior

AWE Atomic Worker Employer

BRH Bureau of Radiological Health

Bq becquerel

DCAS Division of Compensation Analysis and Support 

DCF dose conversion factor 

DHEW U.S. Department of Health, Education, and Welfare 

DOE U.S. Department of Energy 

dpm disintegrations per minute

dpm/m2 disintegrations per minute per square meter 

dpm/m3 disintegrations per minute per cubic meter

DR dose reconstruction 

FDA U.S. Food and Drug Administration 

FUSRAP Formerly Utilized Sites Remedial Action Program

GM geometric mean

GSD geometric standard deviation 

H effective dose

Hp(10) personal dose equivalent 

HT organ dose equivalent (for organ T) 

HASL Health and Safety Laboratory

ICRP International Commission on Radiological Protection

IREP Interactive RadioEpidemiological Program

keV kiloelectron volt 

LOD limit of detection

MCNPX Monte Carlo N-Particle eXtended

MeV mega-electron volts 

μg microgram

mrad millirad

mrem millirem

mrad/hr millirad per hour 
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mR milliroentgen

mR/hr milliroentgen per hour 

n/cm2 neutron per square centimeter

NIOSH National Institute for Occupational Safety and Health

Np neptunium 

OCAS Office of Compensation Analysis and Support 

ORAUT Oak Ridge Associated Universities Team

OTIB Oak Ridge Associated Universities Team Technical Information Bulletin 

pCi picocurie 

pCi/mg picocurie per milligram

pCi/μg picocurie per microgram

PEP program evaluation plan

PER program evaluation report 

PoBe polonium beryllium (neutron source) 

POC probability of causation 

pSv-cm2/n pico-Sievert-square centimeter per neutron 

Pu plutonium 

RU recycled uranium

SCDRR Subcommittee for Dose Reconstruction Reviews 

SRDB Site Research Database

Sv sievert

Sv/Bq sievert per becquerel

TBD technical basis document

Th thorium 

TIB technical information bulletin

U uranium 

XRD x-ray diffraction analysis
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1.0 STATEMENT OF PURPOSE

The National Institute for Occupational Safety and Health (NIOSH) and the Oak Ridge 

Associated Universities Team (ORAUT) have assembled a large body of guidance documents, 

workbooks, computer codes, and other tools to support dose reconstruction (DR). In recognition 

that these supporting elements may be subject to programmatic revisions, provisions exist for 

evaluating the effect of such revisions on the outcome of previously completed DRs. Revisions 

may be prompted by new information, misinterpretation of guidance, policy changes, and/or 

programmatic improvements. 

The process for evaluating potential impacts of programmatic changes on previously completed 

DRs appears in the procedure OCAS-PR-008, Preparation of Program Evaluation Reports and 

Program Evaluation Plans, Revision 2 (NIOSH 2006a). This procedure describes the format and 

methodology to be employed in preparing a program evaluation report (PER) and a program 

evaluation plan (PEP). A PEP describes plans for evaluating specific program details or issues.

A PER provides a critical evaluation of the effect(s) that a given issue/programmatic change may

have on previously completed DRs. This includes a qualitative and quantitative assessment of 

potential impacts. Most important in this assessment is the potential impact(s) on the probability 

of causation (POC) of previously completed DRs with POCs of <50%; i.e., do the changes 

increase a claimant’s POC above 50% so that he or she may be compensated?

During the 115th full meeting of the Advisory Board on Radiation and Worker Health (the 

Advisory Board) held on January 25, 2017, SC&A was directed to perform a review of DCASPER-061, Bridgeport Brass Company, Revision 0 (NIOSH 2015, hereafter referred to as 

PER-061). 

SC&A performs five subtasks in a PER review, each of which is discussed in this report: 

Subtask 1: Assess NIOSH’s evaluation/characterization of the issue and its potential impacts on 

DR to ensure that the issue is fully understood and characterized in the PER. 

Subtask 2: Assess NIOSH’s specific methods for corrective action. In instances where the PER 

involves a technical issue that is supported by document(s) (e.g., white papers, technical 

information bulletins [TIBs], procedures) that have not yet been subjected to a formal 

SC&A review, Subtask 2 will include a review of the scientific basis and/or sources of 

information to ensure the credibility of the corrective action and its consistency with 

current/consensus science. Conversely, if such technical documentation has been 

formalized and previously subjected to a review by SC&A, Subtask 2 will simply provide 

a brief summary/conclusion of this review process. In this particular PER review, 

Subtask 2 will also serve as a site profile review for ORAUT-TKBS-0030, An Exposure 

Matrix for the Adrian Facility and Bridgeport Brass Company, Revision 02 (NIOSH 

2013b, hereafter referred to as TKBS-0030, Revision 02), because it was never reviewed 

by SC&A in its entirety. 

Subtask 3: Evaluate the PER’s approach for identifying the potentially affected DRs and assess 

the criteria by which NIOSH selected a subset of potentially affected DRs for 
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reevaluation. The second step may have important implications in instances where the 

number of previously denied DRs is very large and, for reasons of practicality, NIOSH’s 

reevaluation is confined to a subset of DRs that, based on their scientific judgment, have 

the potential to be significantly affected by the PER. SC&A will also evaluate the 

timeliness for the completion of the PER. 

Subtask 4: Conduct audits of DRs affected by the PER under review. The number of DRs 

selected for audit for a given PER will vary. It is assumed that the selection of the DRs 

and the number of DR audits per will be made by the Advisory Board using case 

selection criteria recommended by SC&A based on the results of Subtask 2. 

Subtask 5: Prepare a written report that contains the results of DR audits under Subtask 4 and

SC&A’s review conclusions. 
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2.0 RELEVANT BACKGROUND INFORMATION 

SC&A performed a focused review of the technical basis document (TBD) ORAUT-TKBS0030, An Exposure Matrix for Bridgeport Brass: Havens Laboratory and Adrian Plant, 

Revision 00, also referred to as the site profile or exposure matrix (NIOSH 2005, hereafter 

referred to as TKBS-0030, Revision 00) in May 2008 (SC&A 2008). Unlike most site profile 

reviews performed by SC&A on behalf of the Advisory Board, this particular review was 

authorized by the Subcommittee for Dose Reconstruction Reviews (SCDRR) as a special 

focused investigation because, at that time, comprehensive site profile reviews were performed 

only for U.S. Department of Energy (DOE) facilities and not Atomic Weapons Employer (AWE) 

facilities. Rather, TBDs for AWE facilities were reviewed only to the extent required to evaluate 

a given DR. The chairman of the SCDRR judged that it was time to begin the process of 

performing more complete reviews of AWE TBDs, starting with Bridgeport Brass. 

Administratively, the review of the TBD for Bridgeport Brass was delivered to the Advisory 

Board as Attachment 1 to the 8th set of DR audit reports, and not as a standalone site profile 

review authorized by a designated site profile review work group. 

SC&A’s focused review of the Bridgeport Brass TBD in 2008 (SC&A 2008) contained five

findings. Discussions of these findings, including NIOSH’s responses, occurred at an SCDRR

meeting on March 12, 2009; SC&A evaluated NIOSH’s responses and discussed the issues 

involved in its white paper, SC&A Follow-Up to NIOSH’s Responses to Bridgeport Brass Site 

Profile Review Findings (SC&A 2009). These issues were again raised as a subject of discussion 

during the February 4, 2013, meeting of the SCDRR. As a follow-up, SC&A issued a special 

memo, Questions/Status of Bridgeport Brass Findings (SC&A 2013), explicitly addressing 

Finding 4, which deals with the statistical issue referred to as “leave-one-out.” SC&A concluded 

that this finding can be withdrawn in recognition of NIOSH adopting a revised method for 

building a co-worker model based on ORAUT-OTIB-0020, Use of Coworker Dosimetry Data 

for External Dose Assignment, Revision 03 (NIOSH 2011c, hereafter referred to as OTIB-0020). 

After these meetings and the publication of the associated technical reports and memo cited 

above, which were based on TKBS-0030, Revision 00, NIOSH issued ORAUT-TKBS-0030, An 

Exposure Matrix for Bridgeport Brass: Havens Laboratory and Adrian Plant, Revision 01 

(NIOSH 2013a, hereafter referred to as TKBS-0030, Revision 01), and then TKBS-0030, 

Revision 02 (NIOSH 2013b) in 2013. NIOSH then issued PER-061 (NIOSH 2015), evaluating

the effect of Revision 02 to the TBD on all previously completed claims, including those first 

made in Revision 01. In recognition of this history, this SC&A PER review includes: 

(1) An assessment of the degree to which TKBS-0030, Revision 02, addresses SC&A’s 

original five findings associated with its review of TKBS-0030, Revision 00, and whether 

any of the findings can be withdrawn based on a reconsideration of these findings, or 

whether any were subsequently resolved in other venues. 

(2) Because SC&A’s review of TKBS-0030, Revision 00, was a focused review, it was 

limited to specific subject areas and did not address some of the topics included in the 

TBD, such as exposures to thorium and recycled uranium (RU), and the residual period 

after operations had ceased. This report includes a summary description of all the 
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technical areas addressed in TKBS-0030, Revision 00, to establish a baseline against 

which to compare and address the changes made to the TBD in subsequent revisions. 

(3) A technical review of new material contained in TKBS-0030, Revisions 01 and 02 

(i.e., a conventional site profile review), including a discussion of the changes from 

TKBS-0030, Revision 00.
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3.0 SUBTASK 1: IDENTIFY THE CIRCUMSTANCES THAT

NECESSITATED THE NEED FOR DCAS-PER-061

As explained in the “Publication Record” section of TKBS-0030, Revision 02 (NIOSH 2013b):

Revision [01 was] initiated to address comments from the Advisory Board 

Subcommittee on Dose Reconstruction. Updated external unmonitored dose in 

accordance with ORAUT-OTIB-0020 [NIOSH 2011c]. Added an internal 

approach to account for remediation efforts in 1976, 1985, and 1995. 

Incorporates formal internal and NIOSH review comments. 

and

Revision [02 was] initiated to address change made by the Division of Energy 

Employees Occupational Illness Compensation to the DOE facilities website 

impacting the Bridgeport Brass facility description and time period. Incorporates 

formal internal and NIOSH review comments. 

These descriptions are very general. More detailed descriptions and assessments of the revisions 

are provided in Section 4.0 of this report. 
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4.0 SUBTASK 2: ASSESS NIOSH’S SPECIFIC METHODS FOR 

CORRECTIVE ACTION

The procedure used in this section to assess NIOSH’s methods for corrective action is to first 

describe the methods used by NIOSH in TKBS-0030, Revision 00, and SC&A’s focused review 

of the TBD (SC&A 2008), and then compare that material to the DR protocols in TKBS-0030, 

Revision 02. This comparison is then followed by a technical review of the methods, data, and 

assumptions adopted in TKBS-0030, Revision 02, for performing DRs. This review will result in 

either a confirmation that the methods are scientifically sound and claimant favorable, or that 

there are some technical deficiencies that require resolution by the work group and the Advisory 

Board. As noted below, SC&A has only one finding and several observations.

As a preface to the review, SC&A found that no substantive changes were made to the dates and 

nature of AWE operations at the Havens Laboratory or the Adrian Plant that appear in the 

Bridgeport Brass TBDs. However, some changes were made to the methods used to reconstruct 

internal, external, and occupational medical exposures during AWE operations and exposures 

during the residual period.

4.1 INTERNAL DOSE RECONSTRUCTION DURING OPERATIONS

4.1.1 Revision 00 to ORAUT-TKBS-0030

As described in TKBS-0030, Revision 00, (NIOSH 2005) and SC&A’s focused review (SC&A 

2008), air monitoring and bioassay data are available for use in DR for many workers at 

Bridgeport Brass. These workers may have been involved in a broad range of activities 

associated with the extrusion operations and handling of rods that generated fumes and oxide 

particulates of natural, slightly enriched, and RU, as well as thorium. These data, along with 

process knowledge, such as uranium enrichment levels, the quantities and types of RU, 

computer-assisted telephone interviews, and job categories, were used to reconstruct the internal 

doses to workers at Bridgeport Brass.

Uranium bioassay data, expressed in units of milligrams per liter of urine, were used as the 

starting point for reconstructing the internal dose and for building a co-worker model for workers 

who had limited or no bioassay data. TKBS-0030, Revision 00, converted the observed mass 

concentrations in urine to activity concentrations using the bounding assumption that the 

uranium had an enrichment level of 2% (i.e., 1,616 picocuries (pCi) uranium (U) per mg, as 

opposed to 683 pCi/mg for natural uranium). The TBD also used the conventional and claimantfavorable assumptions that all the uranium was U-234 and chose the form of uranium that is 

limiting for the organ of concern (i.e., Type S for lung dose and Type M for other organs and 

tissues). 

Because of limitations in the completeness of the uranium bioassay data, NIOSH elected to pool 

the bioassay data and select an upper-end uranium mass concentration in urine to derive the 

annual intake rates for Type S and Type M uranium for the Havens Lab and Adrian Plant. These 

values are provided in Table 3-9 of TKBS-0030, Revision 00. SC&A reviewed the available 

bioassay data and found the values chosen by NIOSH to be reasonably bounding. SC&A also 

compared the derived default uranium intake rate based on bioassay data with air sampling data 
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and found that the air sampling data were compatible with the selected default intake rates based 

on bioassay data. 

Notwithstanding these generally favorable findings, SC&A’s review of TKBS-0030, 

Revision 00, identified one area of concern dealing with limitations in the amount bioassay data 

associated with the pre-1960 time period. Our review revealed that the bioassay data set heavily 

emphasized the post-1960 time period and may not be representative of the pre-1960 time 

period, especially considering that the historical record indicates continuing problems with dust 

accumulations at the Adrian Plant in areas where uranium was extruded through 1956; these 

problems were subsequently corrected. In its review of Revision 00 to the TBD, SC&A 

acknowledged that the importance of this issue was somewhat lessened by the conservative 

assumption that all intakes assumed 2% enriched uranium, when many exposures appear to have 

been to natural uranium. SC&A, therefore, had the following finding: “The site profile would 

benefit from additional analyses that demonstrate that the default intake rates adopted in the 

exposure matrix are claimant favorable for early operational time periods and job categories” 

(SC&A 2008, Finding 1, page 5).

Upon further consideration, SC&A would like to withdraw this finding, as the use of the upper 

95th percentile of the bioassay data, coupled with the assumption that all internal exposures 

assumed 2% enriched uranium, provide a level of assurance that the assigned internal exposures 

during the early years, though based on a relatively limited amount of bioassay data, remain 

claimant favorable because of the use of these bounding assumptions. This is a subjective 

judgment made by SC&A and is certainly an appropriate topic for discussion with the 

subcommittee.

SC&A’s examination of and commentary on TKBS-0030, Revision 00, was the result of a 

focused review performed at the request of the subcommittee; i.e., SC&A limited its review of 

internal exposures to uranium inhalation. However, the TBD also addressed distinctions between 

the Adrian and Havens plants, thorium intakes, RU, and the residual period, which were not 

addressed in SC&A’s focused review. These subjects are covered in the following sections of 

this report. 

In order to derive intakes of radionuclides other than uranium, Revision 00 to TKBS-0030 uses 

the activity fractions shown in Table 3-8 (reproduced here as Table 1). 

Table 1. Reproduction of Table 3-8, “Assumed Activity Fractions of Other Radionuclides 

Relative to Uranium (pCi other radionuclides per pCi uranium)” (NIOSH 2005) 

Uranium Th-228 Th-232 Pu-239 Np-237

1 0.0161 0.0161 0.00246 0.00182

The RU activity fractions for plutonium-239 (Pu-239) and neptunium-237 (Np-237) are based on 

Hanford RU data, and the thorium-232 (Th-232) and Th-238 fractions are derived from the 

assumption that the Th-232 mass intake was 10% of the uranium mass intake based on the mass 

throughput of the two radionuclides at Bridgeport Brass. Other progeny in the Th-232 decay 

chain are ignored because of their very small potential to contribute to the dose. These topics are 
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addressed as part of SC&A’s review of the methods adopted for the reconstruction of internal 

exposures in Revision 02 to TKBS-0030, which follows. 

4.1.2 Revision 02 to ORAUT-TKBS-0030 

This section describes the changes that were made in Revision 02 to TKBS-0030 with respect to 

reconstructing internal doses during the operations period, and includes SC&A’s assessment. No

changes were made to the assigned radionuclide intake rates except for the intake of isotopes 

associated with RU. Revision 02 to the TBD adopts the RU concentration of Battelle-TBD-6000, 

Site Profile for Atomic Weapons Employers that Worked Uranium Metals, Revision 01 (NIOSH 

2011a, hereafter referred to as TBD-6000), which provides generic methods and data for AWEs. 

SC&A previously reviewed Revision 01 to TBD-6000 and all issues have been resolved. Hence, 

SC&A concurs with the values used in Revision 02 to TKBS-0030 to derive uranium and RU 

intake rates.

Thorium operations and the default intake rates of Th-232 and its progeny adopted in TKBS0030, Revision 00, were not addressed in SC&A’s focused review (SC&A 2008). SC&A notes 

that NIOSH’s treatment of thorium remains unchanged through Revision 02 to TKBS-0030. 

Information in the historical records for Bridgeport Brass indicates that thorium was processed 

during the 1950s at both the Havens and the Adrian facilities. However, there are no 

environmental or personnel monitoring records pertaining to thorium. The following is an 

excerpt from Section 3.2 of all the revisions to the Bridgeport Brass TBD (NIOSH 2005, 2013a, 

and 2013b): 

It is not clear when thorium work started at Havens Laboratory. The Adrian Plant 

records indicate that extrusion of thorium most likely started on May 25, 1954. 

Records indicate that thorium work might have slowed down or ceased after 

1955, but no inventory records were found after this date. AEC records indicate 

that there was continuing interest in thorium after 1955. It is favorable to 

claimants to assume that thorium processing continued throughout the AEC work 

periods. To date, no records of thorium air monitoring or bioassay for either site 

have been found. 

NIOSH recommends assigning unmonitored internal dose from thorium exposure for the entire 

operation period. In the absence of thorium monitoring data, NIOSH applies the ratio of the 

processed uranium by mass to processed thorium by mass to the derived uranium intakes. SC&A 

reviewed the document, Source and Special Nuclear Materials Accountability Statements

(Dowling 1955), which reports the amount of natural uranium received from July 1, 1953, to 

June 30, 1954, as 10,821 kg and the amount of thorium received as 190 kg. TKBS-0030, 

Revision 02, rounds these values to 11,000 kg and 190 kg, respectively, which result in the 

thorium representing less than 5% of the uranium by mass. For the time July 1, 1954, through 

May 31, 1955, the amount of normal uranium received was 50,473 kg and the amount of thorium 

was 1,570 kg. The TBD states: “To simplify calculations for both Haven Laboratory and Adrian 

Plant and to account for uncertainty in the relative masses of handled uranium and thorium, this 

analysis assumed that the mass of processed thorium was 10% of the mass of processed 

uranium.” SC&A agrees that, for dose reconstruction, the assumption that the mass of processed 

thorium is 10% of the mass of processed uranium is bounding and claimant favorable. 
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All three revisions to the TBD contain the same assumptions regarding the thorium exposure 

period at both facilities. Section 3.4 states: 

At Havens Laboratory, intakes are assumed to be from naturally enriched 

uranium (0.683 pCi/μg). Thorium intakes should be assumed beginning June 26, 

1952. Recycled uranium contaminants are included after 1952. There are two 

periods of operational exposure: 

• November 8, 1950, to December 31, 1950; and 

• June 26, 1952, to August 27, 1962. 

The 1950 intake is assigned to all Havens Laboratory workers, whose covered 

work period overlapped the 1950 intake period. For unmonitored work periods or 

workers, two exposure scenarios for Havens are shown in Table 3-9. The first can 

be used for unmonitored internal exposures that did not include work during the 

period from April 15 to 21, 1961. The second scenario accounts for the higher 

exposures during the period from April 15 to 21, 1961, and should be used for 

unmonitored exposures that overlapped this period. 

At Adrian Plant, intakes are likely to be from natural (0.683 pCi/μg) or low 

enriched (0.973 or 1.616 pCi/μg) uranium and thorium. Only natural uranium 

enrichment is likely before 1960. The default enrichment assumption for 1960 

through 1962 is 2% (1.616 pCi/μg). 

There is one period of operational exposure: May 25, 1954, to December 31, 

1962. 

At Adrian Plant, thorium and recycled uranium contaminant intakes should be 

assumed for the entire operational exposure period. 

The guidance for unmonitored thorium doses from Revision 02 to TKBS-0030 (page 25) is as 

follows: 

To account for unmonitored thorium exposures at Havens Laboratory and Adrian 

Plant, it is assumed that the thorium intake is equal to 10% of the uranium intake 

by mass for the same period. Natural uranium has a lower specific activity than 

enriched uranium, so it is favorable to claimants to assume natural uranium when 

determining the relative activity of thorium. To determine the relative activities of 

uranium to thorium, the specific activity of 232Th is divided by the specific activity 

of natural uranium and multiplied by 10%. This results in a relative 232Th -touranium intake fraction by activity of 0.0161. Further, it is assumed that 232Th is 

in equilibrium with 228Th, so the 228Th to uranium activity fraction is also 0.0161. 

Exposure from 228Ra (half-life of 5.75 years) is assumed to be insignificant 

because the thorium was likely to have been recently produced and because the 

dose conversion factor is small compared to thorium. Thorium intakes are 

summarized in Table 3-8. 
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Revision 00 to TKBS-0030 uses the uranium air concentrations to derive the uranium intakes, 

but in Revisions 01 and 02, the uranium intakes are derived using bioassay data. Table 3-8 of 

both Revisions 01 and 02 (reproduced below as Table 2) lists the estimated thorium intake rates 

based on the uranium intakes. 

Table 2. Reproduction of Table 3-8, “Estimated Thorium Intake Rates Based on Uranium 

Intakes” (NIOSH 2013a and 2013b)

[Location] Start End Intake 

mode Radionuclide Absorption type

Exposure rate

(pCi/d)

Havens 

Plant

11/8/1950 12/31/1950 Inhalation Th-232 & Th-228 M, S 3.33E0

11/8/1950 12/31/1950 Ingestion Th-232 & Th-228 (a) 6.96E-2

6/26/1952 8/27/1962 Inhalation Th-232 & Th-228 M, S 1.19E-1

6/26/1952 8/27/1962 Ingestion Th-232 & Th-228 (a) 2.48E-1

Adrian 

Plant

5/25/1954 12/31/1962 Inhalation Th-232 & Th-228 M, S 1.19E1

5/25/1954 12/31/1962 Ingestion Th-232 & Th-228 (a) 2.48E-1

SC&A finds this approach to reconstructing internal exposures to thorium to be scientifically 

sound and claimant favorable. 

4.2 EXTERNAL DOSE RECONSTRUCTION DURING OPERATIONS

4.2.1 Revision 00 to ORAUT-TKBS-0030 

Because of limited worker-specific records, NIOSH uses pooled 2-week film badge dosimetry 

data from the Adrian Plant and Havens Laboratory as the basis for constructing a co-worker 

model to derive annual external doses to unmonitored workers during AWE operations. The 

specific methodology is described in Section 4 of TKBS-0030, Revision 00(NIOSH 2005). This 

methodology includes the evaluation of existing external radiation dosimetry and the 

extrapolation and application of the results to individuals who were not monitored over the time

periods of interest. Minimal personnel monitoring data are pooled across several time periods 

and analyzed to determine geometric means (GMs) and geometric standard deviations (GSDs). 

The 95th percentile doses are used to estimate the doses of non-monitored workers during the 

time periods in question. 

Using these assumptions for the Adrian Plant, NIOSH derives an external electron dose 

distribution with a GM of 31.2 millirad (mrad) per nominal 2-week period for unmonitored 

workers, with a GSD of 3.83. For photon doses, the GM per 2-week monitoring period is

4.27 millirem (mrem) and the GSD is 3.70. For the Havens Laboratory, the GM photon dose per 

2-week monitoring period is 3.31 mrem and the GSD is 3.57. For the beta measurements at 

Havens, the GM dose per 2-week monitoring period is 6.99 mrad and the GSD is 5.46. 

NIOSH elected to derive annual 95th percentile doses for penetrating and non-penetrating 

external exposures at both facilities and use these as a constant value for reconstructing annual 

external doses. These derived assigned doses are provided in Table 4-1 of Revision 00 to

TKBS-0030. For the Havens Laboratory, typical annual penetrating and non-penetrating doses 

are 0.335 rem/yr and 1.666 rem/yr, respectively. For the Adrian plant, typical penetrating and 

non-penetrating doses are 0.452 rem/yr and 3.558 rem/yr, respectively. 
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In its original review of Revision 00 to TKBS-0030, SC&A expressed some concern that the 

assigned doses for each year from pooled data may not be representative of all unmonitored 

workers and job categories. However, SC&A believes that the selection of the 95th percentile of 

the dose distribution for each year for each facility reduces the importance of this concern. 

In its focused review of Revision 00 to TKBS-0030 (SC&A 2008), SC&A independently derived 

the external doses but limited its analysis to the Adrian Plant. The evaluation determined that 

“default 95th percentile doses adopted by NIOSH for both non-penetrating and penetrating 

radiation are low by about a factor of 2” (SC&A 2008, Finding 2, page 5). SC&A believes the 

approximate two-fold difference has to do with correlation issues, as explained in the report

(SC&A 2008, page 25): 

TKBS-0030 [NIOSH 2005] states that “correlation” was assumed in the Crystal 

Ball simulation. Assumptions regarding correlation are very important to the 

construction of an exposure matrix, because it affects the degree of conservatism 

inherent in the derivation of default upper-end exposures. For example, let us 

assume we have film badge data for a number of time periods, and that the data 

set for each time period has a log-normal distribution. Using these data, it is 

possible to derive one distribution representing the entire dataset. However, there 

are a number of ways this can be accomplished, depending on whether one 

assumes that the exposures experienced by each worker are independent as a 

function or time or that the exposures a given worker experienced in one time 

period are correlated to the exposures that the worker experienced in other time 

periods. Lacking evidence to the contrary, the usual assumption would be that 

most workers performed the same or similar jobs repeatedly, indicating that the 

assumption of correlation is appropriate for conducting the Crystal Ball 

simulation. If “no correlation” is assumed, the high-end default annual exposures 

would be lower, because the simulation would be based on the assumption that 

the exposures associated with each [2-week] film badge change-out period are 

independent. Hence, under the “no correlation assumption,” a worker might 

receive a high-end exposure during one change-out period and a lower-end 

exposure the next. As a result, the spread of the simulated distribution of annual 

doses is markedly reduced if non-correlation among the doses for each changeout period is assumed when building an exposure matrix. As is demonstrated 

below [please see SC&A 2008 for more detailed discussion of this issue], though 

the site profile states that correlation was assumed, our independent derivation of 

the dose distributions revealed that correlation was not assumed. As a result, we 

believe that the upper-end doses might have been underestimated by a factor of 

about 2. 

In addition to issues associated with correlation, SC&A 2008 found that “exposures to localized 

parts of the body, such as the hands and forearms, from non-penetrating radiation for some 

workers could be missed by a film badge monitoring program and, as a result, the exposure 

matrix may not be claimant favorable for some workers” (SC&A 2008, Finding 3, page 6). This 

conclusion arises from the fact that the non-penetrating dose assignments are based on film 

badge data and do not take into consideration situations where workers’ hands and forearms, and 

perhaps other parts of the body, were in close proximity to a beta source. Under these 
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circumstances, the doses to skin in close proximity to a beta source (such as uranium) may be 

substantively underestimated when based on film badge data. 

Although neutron exposures are included in Section 4.0 of Revision 00 to TKBS-0030, SC&A’s 

review of that document did not address the subject. The TBD (page 27) states: 

Neutron doses were reported for periods ending May 1, 1960 through 

September 4, 1960 and from December 25, 1960 through September 17, 1961 

(except the week ending March 5, 1961). It was noted that calculated neutron 

doses were based on the assumption of a fast neutron source term, and that 1 rem 

equaled 14E6 neutrons/cm2. In the early periods, results were reported as less 

than 0.8E6 neutrons/cm2, which using the conversion above, is consistent with the 

60 mrem reporting threshold. Neutron dosimeters had an unshielded portion and 

a cadmium-shielded portion. All 938 reported results for the shielded portion of 

the dosimeter were less than 60 mrem. Five of the 938 results for the unshielded 

portion of the dosimeter equaled or exceeded the detection threshold (one other 

result reported neither as nonzero or “less than” was 0.3E6 neutrons/cm2

), and 

the maximum result was 1E6 neutrons/cm2, which would equal about 100 mrem. 

Neutron dosimeters were calibrated with a polonium beryllium (PoBe) source. 

This analysis concludes that the reported neutron dose results are consistent with 

the assumption of no significant neutron exposures and the 0.5% rate of positive 

results is not necessarily indicative of workplace neutron exposures.

4.2.2 Revision 02 to ORAUT-TKBS-0030 

Gamma-Beta Exposure 

Although Revision 02 to TKBS-0030 (NIOSH 2013b) uses the same original film badge data 

from the Adrian Plant and Havens Laboratory as previous TBD revisions, the methods used to 

develop a co-worker model were substantially revised using the methodology of OTIB-0020. 

Table 4-1 of Revision 02 to the TBD (reproduced as Table 3 below) provides the external dose 

co-worker doses, as follows: 

Table 3. Reproduction of Table 4-1, “External Dose (rem) for Unmonitored Workers” 

(NIOSH 2013b)

Percentile

Havens Adrian

Gamma Beta Gamma Beta

50th 0.520 0.798 0.596 1.495

95th 1.225 2.932 1.221 5.832

As described above, for the Havens Laboratory, typical annual penetrating and non-penetrating 

doses adopted at the 95% confidence level in Revision 00 to TKBS-0030 are 0.335 rem/yr and 

1.666 rem/yr respectively. For the Adrian plant, 95% confidence level doses for penetrating and 

non-penetrating doses are typically 0.452 rem/yr and 3.558 rem/yr, respectively. As may be 

noted, the assigned doses at the 95% confidence level in Revision 02 are several times higher 
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than those in Revision 00. In addition, Revision 02 to the TBD explicitly addresses extremity 

doses, stating the following (page 29): “If applicable, adjustments to the whole-body dose should 

be made for the extremities (e.g., hand and forearms) to account for geometry issues using the 

guidance in DCAS-TIB-0013, Selected Geometric Exposure Scenario Considerations for 

External Dose Reconstruction at Uranium Facilities (NIOSH 2010...).” 

TKBS-0030, Revision 02, assumes that, for the purpose of estimating doses from external 

exposure to uranium, most photons have energies in the 30–250 kiloelectron volt (keV) range. 

While this statement is correct with respect to photon fluence, we note that 87% of the photon 

energy flux is from photons with energies >250 keV (see below). Because absorbed dose is 

expressed in terms of deposited energy, the energy flux is the more relevant metric. 

In conjunction with a scoping analysis to estimate the relative neutron dose from a representative 

external exposure scenario described below, we used Monte Carlo N-Particle eXtended 

(MCNPX) to calculate the personal dose equivalent, Hp(10), at distances of 1 foot and 1 meter

from a uranium rod, 77 inches long and 1.405 inches in diameter. These are the average 

dimensions of the uranium rods received from Fernald at the Havens Laboratory in 1954, after 

cutting to the desired lengths. The MCNPX analysis shows that an average of 90% of the doses 

are from photons with Eγ > 250 keV. 

Although TKBS-0030, Revision 02, acknowledges that the photon radiation from uranium metal 

would have a hardened spectrum, it maintains that assigning the doses to photons in the 30–250 

keV range is claimant favorable. A review of the factors for converting Hp(10) to organ dose 

listed in OCAS-IG-001, External Dose Reconstruction Implementation Guideline, Revision 3 

(OCAS 2007), shows that for 13 of the 17 organs listed, the DCFs for the anterior-posterior (AP) 

orientation are higher for photons with Eγ > 250 keV. 

Observation: SC&A recommends that NIOSH assume photon energies >250 keV, except 

when calculating doses to the surface of bone, the testes, the thymus, and the thyroid.

Based on this discussion and comparison of the assigned external penetrating and nonpenetrating doses in Revision 00 versus Revision 02 to TKBS-0030, we find that, with one 

exception, the external penetrating and non-penetrating doses assigned to workers at Bridgeport 

Brass during AWE operations in Revision 02 to the TBD are scientifically sound and claimant 

favorable. In addition, SC&A and the Advisory Board previously reviewed OTIB-0020 (NIOSH 

2011c), the version used in support of Revision 02 to the TBD, and all issues are resolved. 

The one exception has to do with the assumption that the hardened photon spectrum can be 

assumed to have an energy spectrum of 30–-250 keV. As discussed above, we believe a more 

scientifically and claimant-favorable approach for most organs would be to assume the photon 

energy spectrum to be >250 keV. 

Exposure to X-Ray Crystallography

The Havens Laboratory performed x-ray crystallography (also known as x-ray diffraction 

analysis or XRD) on uranium and thorium samples. According to TKBS-0030, Revision 02 

(page 13): 
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There is mention of X-ray crystallography work in the AEC contract to inspect 

metal samples, but no information about the design or safety precautions of this 

analytical equipment was found. Late 1958 to 1960 biweekly film badge results 

for areas specified as “X-ray” were usually reported as <10 mrem. 

TKBS-0030, Revision 02, does not further discuss this radiation source. 

Although the TBD assigns photon doses based on film badge dosimetry reports, using either data 

for individual workers or a coworker model in cases where there are no individual dosimetry 

data, such readings cannot be used to determine exposures to XRD. Because of the small 

diameter of secondary x-ray beams from the XRD apparatus, not all radiation received by a 

worker would be registered on his film badge. Portions of the worker’s body, especially the skin, 

fingers, and upper extremities, could have been exposed to scattered radiation that was not 

measured by the dosimeters. According to Lubenau et al. (1969, page 741): 

Personnel monitoring devices, such as film badges, are local radiation recording 

devices. Results of a film badge worn on the shirt pocket may not be 

representative of the exposures to the eye or to the hands. Energy response of the 

devices must be considered.

Exposure to XRD apparatus was a recognized hazard, as evidenced by the fact that the Bureau of 

Radiological Health (BRH) sponsored a conference on this topic (Moore et al. 1971). The 

participants stressed the need for improved monitoring and radiation safety. According to 

Rudman (1971, page 72): “A number of manufacturers have recently (within the last five years) 

marketed special shutter assemblies that include various fail-safe features. However, there are 

still very many older x-ray units in operation and these must be checked very carefully.” This 

need would have been even more acute 10 years earlier—the time of the XRD work at 

Bridgeport Brass.

In the case of the Carborundum Company, another AWE site, SC&A identified a former worker 

who had performed XRD operations during the covered period at that site. This worker provided 

sufficiently detailed information about use of the XRD apparatus that it was possible to construct 

an exposure scenario that bounded the radiation exposures from this source. NIOSH should 

attempt to identify former workers or other data sources that would enable the construction of a 

bounding x-ray crystallography exposure scenario for Bridgeport Brass. 

Neutron Exposure

Sections 4.0 of TKBS-0030, Revisions 01 and 02, include the same discussion of neutron doses 

as Revision 00, with only minor editing changes; the discussion is reproduced in the preceding 

section of this report. SC&A notes that a letter of March 9, 1961 (Grella 1961), from Bridgeport 

Brass to the Atomic Energy Commission (AEC) reports that two types of film badge holders and 

film inserts were supplied to the Adrian, Michigan, plant by Controls for Radiation, Inc.: a betagamma badge and one that also included a neutron filter (strip of cadmium) and a criticality 

monitoring foil (strip on indium). The Bridgeport, Connecticut, plant badges just recorded betagamma radiations (Grella 1961). 
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Another letter, dated January 16, 1963 (Jefferson 1963), from Bridgeport Brass to the AEC 

summarizes the total number of badges used in fiscal year 1962 at two sites: The Havens Lab 

(which had moved to Seymour, Connecticut) used 190 beta-gamma badges; and the Adrian, 

Minnesota, plant (which had moved to Ashtabula, Ohio) used 460 beta-gamma badges and 260 

neutron badges (Jefferson 1963). This information about the badging practices supports the 

conclusion that at least some of the Bridgeport Brass employees were monitored for neutron 

exposures some of the time. 

SC&A wanted to investigate several TBD claims related to neutron monitoring but was 

hampered by the TBD (in the previously quoted paragraph) not citing sources for neutron dose 

reports (938 results), the distribution of dose values, and the conversion factor for fast neutrons 

of 1 rem equaling 1.4×106 neutrons/cm2. SC&A was unable to locate the data sources in the Site 

Research Database (SRDB). Nonetheless, SC&A’s understanding of neutron doses associated 

with plants that worked with natural or low-enriched uranium metal and oxides (natural uranium 

in the Havens plant and up to 2% enrichment in the Adrian plant, as noted in AEC 1960) and 

their short-lived progeny supports the TBD’s conclusion that the potential penetrating neutron 

dose (from spontaneous fission and alpha-n reactions) is negligible compared to that from 

gamma radiation. 

SC&A independently derived its own neutron fluence-to-dose conversion factor to compare to 

NIOSH’S claim that 1 rem equals 1.4×107

 neutrons per square centimeter (n/cm2) and found that 

the claim is supported. Table A.5 of ICRP Publication 116, Conversion Factors for Radiological 

Protection Quantities for External Radiation Exposures (ICRP 2010), gives a conversion factor 

for 1 mega-electron volt (MeV) monoenergetic neutrons with AP geometry (i.e., the subject is 

facing the incoming neutrons) as 301 pico-sievert-cm2/neutron (pSv-cm2/n). Converting units to 

allow comparison to the NIOSH number: 

(301 pSv-cm2/n) × (100 rem/Sv) × (1.00×10-12 rem/pico-rem) = 3.01×10-8 rem-cm2/n. 

Inverting this number yields 1.4×107 n/cm2 per rem, which is the value used by NIOSH. 

This is the conversion factor for effective dose (often denoted as H), which must be converted to 

organ dose equivalent (often denoted as HT, where T represents a particular tissue or organ) to 

determine damage to that organ. Appendix A to OCAS-IG-001 (NIOSH 2007) provides 

extensive neutron DCFs for many organs in units of rem/neutron per cm2

. For example, the 

appendix gives a value of 3.238×10-8 rem to the stomach per neutron/cm2 for AP geometry for 

neutrons with energies ranging from 0.1 to 2 MeV. This is consistent with the values in ICRP 

Publication 116 (ICRP 2010). 

Section 4.0 of the site profile states that the maximum neutron dosimeter reading of the 938 

values over the reported period corresponds to a fluence of 1×106 n/cm2, or a dose of about 

100 mrem per 2-week change-out. SC&A acknowledges that NIOSH could have assigned a 

bounding value of annual neutron dose of 100 mrem per change-out (i.e., 2-week period) × 26 

change outs- per year, which would equal 2.6 rem/yr. This is not an insignificant annual dose as 

compared to the annual beta and gamma doses described above (i.e., approximately 1 rem/yr 

assigned for photon doses). However, SC&A believes that the potential for neutron exposures of 

this magnitude is implausible at both the Havens Laboratory and the Adrian Plant because there 
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are no reactors, sources of highly enriched uranium, or uranium nitrate or uranium hexafluoride 

(both of which could have been associated with alpha/n reactions). Hence, at these sites, we 

believe that it would be inappropriate to use the bounding approach described above or the 

1⁄2-the-minimum-detection-limit approach, which is often used as a default approach for building 

a co-worker model. In this respect, we concur that the neutron doses at the two facilities were 

likely a very small fraction of the gamma doses.

One final point, although not stated explicitly in TKBS-0030, Revision 02, NIOSH’s claim in 

Section 4.0, page 29, that “the reported neutron dose results are consistent with the assumption 

of no significant neutron exposures and the 0.5% rate of positive results is not necessarily 

indicative of workplace neutron exposures,” leads SC&A to infer that NIOSH assigns a zero 

neutron dose to employees; this assumption should be confirmed or explained by NIOSH. As 

described below, SC&A independently estimated the neutron-to-photon ratio associated with 

spontaneous fission of about 0.01 for uranium metal rods of the types handled at Bridgeport 

Brass. 

TKBS-0030, Revision 02, stated that neutron doses were reported during two periods. Because

few of the results listed doses above the limit of detection (LOD), the TBD concluded that there 

were no significant neutron exposures. To test this hypothesis, SC&A calculated the Hp(10) 

neutron doses, using the same exposure geometry as for the photon dose analysis described 

above. We found that the neutron doses constituted 1.1% of the photon doses at the two dose 

points. Table 4-1 of the TBD lists 95th percentile gamma doses of ~1.2 rem/yr at the two 

Bridgeport Brass facilities. Applying the neutron-to-photon dose ratio resulting from our 

MCNPX analyses, we obtain annual neutron doses of ~12 mrem, which could be significant in 

cases with POCs just under 50%. This result is also consistent with the film badge data, because

such dose rates would not be detected by the biweekly neutron film badges, which were reported 

to have an LOD = 60 mrem. The MCNPX analyses show that ~1% of the neutron doses are from 

neutrons with energies ≤100 keV, 72% are from 100 keV ≤ En ≤ 2 MeV, and 27% from 2 MeV 

≤ En ≤ 20 MeV. 

Finding 1. The site profile should assess the neutron doses from spontaneous fission.

Although the potential neutron doses from spontaneous fission were likely to be extremely 

small, as compared to the assigned photon doses, SC&A believes a quantitative estimate of 

the doses from spontaneous fission should be performed. 

4.3 REMEDIATION/RESIDUAL PERIOD

SC&A’s original focused review of Revision 00 to TKBS-0030 (NIOSH 2005) did not address 

the remediation/residual period. Therefore, this section presents a review of the Section 5 of 

Revision 02 to TKBS-0030 (NIOSH 2013b), which addresses this subject. AWE operations at 

the Havens Laboratory and Adrian Plant ended in 1962, and both facilities were immediately 

surveyed following decontamination and the removal of equipment in 1962. Revision 02 to the 

TBD refers to the Area Contamination Survey Report (JG 1962) as the basis for determining that 

the residual radioactivity in 1962 at both facilities following decontamination was negligible and 

did not contribute substantively to post-AWE exposures. JG 1962 provides a description of the 

results of the surveys and may be used to help confirm that there were, in fact, insignificant 

levels of residual contamination following the termination of AWE operations and the 1962 
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decontamination activities. Both sites were further surveyed under the Formerly Utilized Sites 

Remedial Action Program (FUSRAP) implemented in 1977 and 1985. 

The documents packaged in SRDB Ref. ID 9588 reveal that, following the 1962 

decontamination operations and subsequent surveys, the site met the decontamination criteria at 

that time. It also appears that a definitive characterization of the residual contamination of the 

Havens Laboratory was performed as part of the FUSRAP program, which began in August 

1980. The results of these surveys revealed both beta and gamma levels at most locations within 

the range of natural background (i.e., 5–10 microroentgen per hour (μR/hr) gamma and on the 

order of 0.02 mrad/hr beta). Some spots near walls had residual contamination on the order of 

twice background. However, the report explains that these slightly elevated readings were due to 

the slightly elevated levels of naturally occurring radionuclides in the brick walls, rather than 

from contamination. The maximum alpha contamination observed was at one spot with a level of 

156 disintegrations per minute per 100 square centimeters (dpm/100 cm2

) (ORNL 1985). 

Included among the files packaged in SRDB Ref. ID 9588 is correspondence between the Health 

and Safety Laboratory (HASL) and Bridgeport Brass beginning in 1962. The correspondence 

provides a wealth of information on the contractual relationship between Bridgeport Brass and 

the government related to AWE activities. There is also extensive correspondence between the 

AEC and Bridgeport Brass describing the nature and extent of the AWE activities at the facility 

beginning in 1950 up through 1962. 

Among the correspondence packaged in SRDB Ref. ID 9588 is a memorandum dated 

October 30, 1964, from A.J. Breslin, Director of the HASL, to John W. Ruch, Director of the 

Feed Materials Division of Oak Ridge National Laboratory, “Contamination Survey at Reactive 

Metals, Inc., Seymour, Connecticut” (Breslin 1964). The memorandum states that a radiological 

survey of the facility revealed that the process areas have been satisfactorily cleaned up of 

uranium contamination and can be released for unrestricted use. A detailed description of the 

results of the survey is provided, which is consistent with the summary in Revision 02 to TKBS0030. Some of the results are as follows:

• Direct alpha: 50% <300 dpm/100 cm2

; 87% <2,000 alpha dpm/100 cm2; 97% <200 alpha 

dpm/100 cm2; max 25,000 alpha dpm/100 cm2

• Smear alpha: 20 to 90 dpm/100 cm2

• Highest beta/gamma 0.5 mrad/hr

The Breslin memorandum concludes that the activity levels were quite low. The memorandum 

also states that they were not aware of any standards at that time to which these measurements 

could be compared. However, the memorandum references made to “Health Protection Program 

Review of Special Metals Development Department, Reactive Metals, Inc., Seymour, 

Connecticut - June 1964.” The memorandum includes many drawings where the results of 

individual measurements are provided on sketches of the facility. A draft letter from R.J. Hart, 

Manager, to William E. Sides, Director of Manufacturing of Brass Mill Products, to the manager 

of the facility refers to a May 1976 report by the U.S. Energy Research and Development 

Administration clearing Bridgeport Brass from any radiological concerns (Hart [n.d.]). In 

addition, a letter from Arthur J. Whitman to Andrew Wallo, of The Aerospace Corporation, 
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dated October 23, 1985, states that surveys reveal that Bridgeport Brass can be eliminated from 

the FUSRAP (Whitman 1985). 

Using these data, a crude estimate of the potential exposures can be derived by assuming surface 

contamination of 90 dpm alpha/100 cm2. With the DCFs in Table 3.10 of TBD-6000 (NIOSH 

2011a), the external photon dose associated with this surface alpha contamination is estimated as 

follows: 

External photon exposure: 

(90 dpm/100 cm2) × (3.94×10-10 mR/hr per dpm/m2) × (1.00×104 cm2/m2) = 3.55×10-6 mR/hr

This external photon exposure rate is negligible, even assuming 2,000 hours per year of 

exposure. 

Inhalation exposure: 

(90 dpm/100) × (1×10-6/m) × (1.00×104 cm2/m2) = 9.00×10-3 dpm/m3

(9.00×10-3 dpm/m3

) × (1.2 m3/hr) × (2,000 hr/yr) × (7.5×10-5 Sv/Bq) × (1.00×105 mrem/Sv) × 

(60 Bq/dpm) = 10 mrem/yr

SC&A considers this a negligible dose because the assumptions are quite bounding and the 

resulting dose is very small.

In light of these data and correspondence, and the associated dose calculations, SC&A concurs 

with TKBS-0030, Revision 02, that any potential exposures during the residual period were 

negligible, and that exposures associated with decontamination operations were captured by the 

survey programs and their associated exposure matrix.

4.4 OCCUPATIONAL MEDICAL DOSE

Section 4.1 of Revision 02 to TKBS-0030 (NIOSH 2013b) recommends using the default values 

in ORAUT-OTIB-0006, Dose Reconstruction from Occupational Medical X-Ray Procedures

(NIOSH 2011b), to reconstruct occupational medical dose. This revision to ORAUT-OTIB-0006 

has been reviewed and accepted by the Advisory Board. Accordingly, SC&A has no comments 

on the methods used in the TBD with regard to occupational medical dose. 
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5.0 SUBTASK 3: EVALUATE THE PER’S STATED APPROACH FOR 

IDENTIFYING THE UNIVERSE OF POTENTIALLY AFFECTED 

DOSE RECONSTRUCTIONS

Section 3.0 of PER-061 (NIOSH 2015, page 2) states that:

In order to evaluate the effect of revision 2 of the TBD on previously completed 

claims, a search was conducted for all completed claims that had a probability of 

causation (POC) of less than 50% and employment during the operational period 

at either site. This search identified 14 claims with employment at the Haven’s 

Lab and 36 claims with employment at the Adrian Facility. All 50 claims had 

employment during the operational period for the particular site. 

A new dose estimate was performed for all 50 claims using revision 2 of the TBD 

as well as all applicable approved dose reconstruction methods. The resulting 

probability of causation (POC) was below 45% for 47 claims. The POC was 

greater than 50% for one claim. The remaining 2 claims resulted in a POC 

between 45% and 50%. For those two claims, IREP was run 30 times at 10,000 

iterations per NIOSH procedures. The resulting POC was less than 50% for both 

claims.

SC&A concurs with this comprehensive approach to implementing this PER. 

6.0 SUBTASK 4: CONDUCT AUDITS OF A SAMPLE OF DOSE 

RECONSTRUCTIONS AFFECTED BY DCAS-PER-061

SC&A recommends the selection of a sufficient number of claims where the external, internal, 

and occupational medical DRs can be checked. At a minimum, we recommend three cases with 

any combination of these three exposure pathways. 
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7.0 REFERENCES

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